ML101970339

Robinson Initial Exam 2008-301 Draft RO Written Exam

ML101970339
Person / Time
Site:	Robinson
Issue date:	06/28/2010
From:	Progress Energy Carolinas
To:	NRC/RGN-II
References
50-261/08-301
Download: ML101970339 (421)

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{{#Wiki_filter:HLC-08 NRC Written Exam 1. Given the following: -The plant is operating at 100% RTP. -The PRT is at 74% and 110 of. -A load rejection results in a Reactor Trip. -Following the trip, a Pressurizer Safety valve opens, and will NOT close. -The PRT rupture disks function as designed. -Containment pressure peaks at 35 PSIG. Which ONE (1) of the following is the approximate MAXIMUM temperature indicated by the Safety Valve tailpiece RTD during the entire event? A. 281 of. B. 338 of. C. 547 of. D. 651 of. 1 do {\o+ 5('.G P VO,.\LA-* £, gt'l... 1 Name: ________________ HLC-08 NRC Written Exam Form: 0 Version: 0 1. 000008 AK1.01 OOlIPZR VAPOR SPACE ACCI/1/1/3.2/3.7IRO/HIGH/N/NNEW -200S/THERMO CHAP 3-00S Given the following: -The plant is operating at 100% RTP. -The PRT is at 74% and 110 of. -A load rejection results in a Reactor Trip. Following the trip, a Pressurizer Safety valve opens, and will NOT close. -The PRT rupture disks function as designed. -Containment pressure peaks at 35 PSIG. Which ONE (1) of the following is the approximate MAXIMUM temperature indicated by the Safety Valve tailpiece RTO during the entire event? A. 281 of. B:' 338 of. C. 547 of. D. 651 of. The correct answer is B. A: Incorrect -Saturation temperature for Containment pressure is 35 PSIG. B: Correct -PRT saturation temperature for 100 PSIG. C: Incorrect -Saturation temperature for No-load T AVG. 0: Incorrect -Saturation temperature for 2235 PSIG. Exam Question Number: 1

Reference:

Steam Tables; SO-059 PZRlPRT, Page 12. KA Statement: Knowledge of the operational implications of the following concepts as they apply to a Pressurizer Vapor Space Accident: Thermodynamics and flow characteristics of open or leaking valves. History: New -Written for HLC-08 NRC Exam. Tuesday, June 17,20081:20:59 PM 1 SD-059 PRESSURIZER SYSTEM provides Low Temperature Over pressure Protection (LTOPP) for the RCS when it is water-solid, that is, at temperatures that a steam bubble cannot be maintained in the PZR. The PORVs are pneumatic valves, with nitrogen being supplied by the Plant Nitrogen System. To assure proper operation of the LTOPP System, the Instrument Air System is valved into service as a backup for the Plant Nitrogen System whenever LTOPP is in service. For Appendix R Safe Shutdown purposes, PCV -455C and PCV -456 are classified as Hi I Lo Pressure Interface Valves. During normal operation these valves are closed except as described above. During a postulated Appendix R fire in Fire Area A5, these valves are required to be closed initially in order to maintain the RCS operating parameters (Pressurizer level, pressure and temperature) within the expected ranges. Key operated switches are located on the Containment Fire Protection Panel in the Control Room. These are two position switches for each PORV (Normal or Isolate). The basic function of the Pressurizer PORV Normal I Isolate switches is to de-energize the electrical circuit during a postulated fire in Fire Area A5 to preclude (to the extent possible) the possibility of spurious operation of these valves. The addition of the isolation switches in conjunction with the actions taken in FP-001 ensure that the circuit is de-energized and protected against the possibility of conductor to conductor hot shorts within a cable. This will result in the Pressurizer PORVs failing closed. 3.6 PZR Relief Tank (PRT) Design Pressure Design Temperature Normal Operating Pressure . Normal Operating Temperature Normal Water Volume Rupture Disc Release Pressure Rupture Disc Relief Capacity Internal Volume 100 psig 340°F 3 psig 120°F 900 fe 100 psig 900,000 #/hr saturated steam 1300 fe The PRT is a horizontally mounted 1300 fe tank inside the Containment Vessel (CV). It has a design temperature and pressure of 340°F and 100 psig respectively. It is piped to the PZR safety and PORVs by a 12" line. It is protected from over pressurization by.nYQ. rupture that will relieve pressure to the Containment Vessel at approximately lQ.Q. psig. The rupture discs are designed to pass 900,000 lbs/hr. saturated steam. The discharge from the High Point Vent System can also be directed to the PRT. The PRT also collects leakage and liquid from various system relief valves located inside PZR Page 12 of 27 Revision 9 INFORMATION USE ONLY HLC-08 NRC Written Exam 2. During implementation of EPP-9, TRANSFER TO COLD LEG RECIRCULATION, how is the Charging Header isolated and why? A. IVSW is automatically aligned. To prevent leakage by seating the RCP Seal Injection line check valves. B. A manual valve alignment of IVSW is performed. To prevent leakage by seating the RCP Seal Injection line check valves. C. IVSW is automatically aligned. To prevent leakage from Containment. D. A manual valve alignment of IVSW is performed. To prevent leakage from Containment. 2 "f\" <\"(!.' I Y'\oT CL.,;t;z? ""e",\'-,){-I:V C/Li.Jl 'b\e. I.L><llM..\,I.. X.\I5W cJ-.I;s"", '1'.,,1--1'N?,,.:t-IS ex. ",\CUAS;\'\<l . HLC-08 NRC Written Exam 2. 000011 EA1.12 001ILBLOCAllI1I4.1I4.4/ROILOW/N/A/NEW -200S/EPP-9-005 During implementation of EPP-9, TRANSFER TO COLD LEG RECIRCULATION, how is the Charging Header isolated and why? A. IVSW is automatically aligned. To prevent leakage by seating the RCP Seal Injection line check valves. B. A manual valve alignment of IVSW is performed. To prevent leakage by seating the RCP Seal Injection line check valves. C. IVSW is automatically aligned. To prevent leakage from Containment. Dr A manual valve alignment of IVSW is performed. To prevent leakage from Containment. The correct answer is D. A: Incorrect -During EPP-9 IVSW is manually aligned to the Charging header. IVSW is applied to the Charging Pump discharge side of the check valves. This would open the check valve, NOT seat them. B: Incorrect -During EPP-9 IVSW is manually aligned to the Charging header. IVSW is applied to the Charging Pump discharge side of the check valves. This would open the check valve, NOT seat them. C: Incorrect -During EPP-9 IVSW is manually aligned to the Charging header. IVSW is applied to the Charging Pump lines to preveent leakage if Continament is pressurized above Charging Pump header pressure and the Charging Pumps are stopped. 0: Correct -During EPP-9 IVSW is manually aligned to the Charging header. IVSW is applied to the Charging Pump lines to preveent leakage if Continament is pressurized above Charging Pump header pressure and the Charging Pumps are stopped. Exam Question Number: 2

Reference:

EPP-9, Pages 32, 33; EPP-9 BD Page 37; SD-021 CVCS, Figure 10. KA Statement: Ability to operate and monitor the following as they apply to a Large Break LOCA: Long-term containment of radioactivity. History: New -Written for HLC-08 NRC Exam. 2 EPP-9 TRANSFER TO COLD LEG RECIRCULATION CONTINUOUS USE ATTACHMENT 2 Attachment 2 (Page 2 of 3) Rev. 31 Page 32 of 40 ************************************************************************** CAUTION The Control Room will be initiating CV Sump Recirculation. This may result in high radiation in the Auxiliary Building.

• CVC-297A.

B. and C are located in the Northwest corner above Seal Inj Filter shield wall.

• CVC-293A.

293C. 292A. and 295 are located in Northwest corner outside Seal Inj Filter shield wall.

• CVC-295A is located in Northwest corner above Seal Inj Filters.
• CVC-309A and 202A are located on West wall adjacent to HCV-121. 5. Verify CLOSED the following valves in the Charging Pump Room: a. CVC-297A.

RCP "A" SEAL WATER FLOW CONTROL VALVE. ------? b. CVC-297B. RCP "B" SEAL WATER FLOW CONTROL VALVE. CVC-297C. RCP "c" SEAL WATER FLOW CONTROL VALVE. CVC-293A. SEAL INJECTION FILTER "A" OUTLET. --7 e. CVC-293C. SEAL INJECTION FILTER "B" OUTLET. ?f. CVC-292A. SEAL INJECTION FILTER PIC-157 ISOLATION.

g. CVC-295. SEAL INJECTION FILTER "A" AND "B" BYPASS. h. CVC-295A.

SEAL INJECTION FILTERS OUTLET VENT. i. CVC-309A. HCV-121 BYPASS. j. CVC-202A. HCV-l21 OUTLET. EPP-9 TRANSFER TO COLD LEG RECIRCULATION CONTINUOUS USE ATTACHMENT 2 Attachment 2 (Page 3 of 3) Rev. Page

6. Open the following valves at the IVSW Tank Area Manual Header: -a. IVSW-16, IVSW TO PEN 24, CHARGING LINE ISOLATION.

b. IVSW-16A, IVSW TO PEN 25, 26, & 27, RCPS SEAL INJECTION.

7. Notify Control Room that Attachment 2 is complete.

-END -31 33 of 40 RNP WOG BASIS/DIFFERENCES STEP STEP 3 4 1N5, 2N5 3N5, 4N5 5 6 7 I EPP-9-BD RNP BASIS Closing CVC-282 isolates the charging line outside CV penetration P-24. This is necessary because the Charging Pumps will be stopped when the RWST is emptying <<27%) which depressurizes the header below design CV pressure. The isolation of the Charging and Seal Injection lines and initiation of IVSW is a corrective action for SCR 90-084 and is documented under correspondence RNPD/91-0275. The reason the penetrations are isolated is that, although IVSW is not credited in the accident analysis, the Appendix J leakage testing program requires all valves to be tested at RFO intervals. If the valves are sealed by a qualified seal system (such as IVSW), the testing may be performed with water and the leakage excluded from the overall leakage rate of the CV. The RNP testing program is set up to take advantage of this portion of the Appendix J testing requirements. Since the valves are tested under these requirements, the valves must be closed and IVSW must be initiated when the system is depressurized. An alternative would be to leave at least one Charging Pump in service during the duration of the accident. This however would create water inventory problems in the CV. RNP BASIS The operator must notify the Control Room when the critical steps of the attachment are complete in order to expedite the restoration of core cooling. RNP BASIS Location of valves is shown to assist in performing the actions. They are included as a note to reduce clutter in the action steps. RNP BASIS Charging flow isolation to the RV was accomplished by cloSing CVC-282 in a previous step. This step completes the isolation of charging by closing manual isolation valves for RCP seal water and seal injection filters. In addition, the HCV-121 outlet isolation is closed and its bypass is verified closed. RNP BASIS The IVSW System is activated to enhance CV isolation at penetrations 24, 25, 26 and 27. RNP BASIS This step directs the operator to notify the control room that this attachment is complete. This is a standard step at the end of attachments. Attachment 3 The following steps are contained in Attachment

3. There are no equivalent ERG steps, accordingly, there are no step differences.

These steps are necessary to increase CV Sump PH for long term corrosion control per Westinghouse NSAL 93-016. This letter describes the long term corrosion control concerns if a SBLOCA occurs which does not result in a Spray Actuation. This attachment aligns RHR and SI in a manner to establish a suction path for the CV Spray Pumps (Piggy-Back Mode). This alignment is performed in the same manner as that done in the main body of the procedure if CV Spray is required. Rev. 31 Page 37 of 521 fSEAl3-1 L ____ J fSW21 '--___ .J r----, I SEALI I L ___ J SEAL I RETURN SEAL INJECTION cvcs-FIGURE-l 0 RCPC SEAL I RETURN SEAL I BYPASS fROM RCPB&C CVC-307 HVC-137 FROM LOOP

• I RCDT LEG LOOP 2 EXCESSlETDOVVN 389 I HEAT EXCHANGER I --SEAL WATER HEAT EX. TO VCT I

-bf1 ---' ,:>-3 7'1 { SEAL WATER INJECTION filTERS --S tt elI--r TORCPB&C (rof"" xvSvJ INFORMATION USE ONLY FROM CHARGING PUMP DISCH. IcvcsPfOJ HLC-08 NRC Written Exam 3. Given the following: -The plant is operating at 8% RTP making preparations for synchronizing to the grid. -At 0800, Annunciator APP-001-D2, Rep #1 SEAL LEAKOFF HI FLOW, alarms. -At 0802, the RO reports #1 sealleakoff is 5.2 GPM for Rep 'A'. -At 0805, the RO reports #1 sealleakoff is now at 5.7 GPM and increasing for Rep 'A', and #1 sealleakoff for Rep '8' and Rep 'e' have decreased from 2.4 GPM to 1.1 GPM. -APp-001-e5, Rep STANDPIPE HIILO LVL alarm is illuminated. -Status light indicates A LOW. Which ONE (1) of the following describes the Rep seal failure and the action required? A. #1 seal has failed; trip the reactor, trip Rep "A"; enter PATH-1. B. #2 seal has failed; trip the reactor, trip Rep "A"; enter PATH-1. C. #1 seal has failed; place plant in MODE 3 lAW GP-006, and trip Rep "A". D. #2 seal has failed; place plant in MODE 3 lAW GP-006, and trip Rep "A". ,\ ,. \ <, r,.t '\-<0 '\V'QV--e CuIISCv<{!' c,\r--oices . A, -!), I W Qi\, --r;..iv \("",,1T,. B, -IF L ,--r;:'1> ... C, -LI ,5ID:LAWG-P-&. t>. 11=--2-, ." 'D ::r;m.J r;.f-t,. 3 HLC-08 NRC Written Exam 3.000015 AK2.10 OOlIRCP MALFUNCTIONSI1I1I2.8/2.8IROIHIGHINIA/ROBINSON -2002/AOP-018-004 Given the following: -The plant is operating at 8% RTP making preparations for synchronizing to the grid. -At 0800, Annunciator APP-001-D2, RCP #1 SEAL LEAKOFF HI FLOW, alarms. -At 0802, the RO reports #1 sealleakoff is 5.2 GPM for RCP 'A'. -At 0805, the RO reports #1 sealleakoff is now at 5.7 GPM and increasing for RCP 'A', and #1 seal leakoff for RCP 'B' and RCP 'c' have decreased from 2.4 G PM to 1.1 G PM. -APP-001-C5, RCP STANDPIPE HIILO LVL alarm is illuminated. -Status light indicates A LOW. Which ONE (1) of the following describes the RCP seal failure and the action required? A'I #1 seal has failed; trip the reactor, trip RCP "A"; enter PATH-1. B. #2 seal has failed; trip the reactor, trip RCP "A"; enter PATH-1. C. #1 seal has failed; place plant in MODE 3 lAW GP-006, and trip RCP "A". D. #2 seal has failed; place plant in MODE 3 lAW GP-006, and trip RCP "A". The correct answer is A. A: Correct -Indications for #1 seal failure and plant is in MODE 1. B: Incorrect -All indications support #1 Seal failure, not #2. C: Incorrect -With plant in MODE 1, required actions is to trip the Reactor instead of entering GP-006. D: Incorrect -All indications support #1 Seal failure, not #2. With a #2 seal failure and the plant in MODE 1, correct actions are to enter GP-006. Exam Question Number: 3

Reference:

APP-001-C5 and D2; AOP-018, Pages 5, 10; SD-001, RCS, Figures 22, 23 and 25. KA Statement: Knowledge of the interrelations between the Reactor Coolant Pump Malfunctions (Loss of RC Flow) and the following: RCP indicators and controls. History: Tuesday, June 17, 2008 1 :21 :00 PM 3 APP-001-02 ALARM Rep #1 SEAL LEAKOFF HI FLOW *** WILL REFLASH *** AUTOMATIC ACTIONS 1. None Applicable CAUSE 1. Failure of RCP Number 1 Seal OBSERVATIONS

1. Labyrinth Seal ilP (PI-125A, PI-128A, PI-131A) 2. RCP Seal Leakoff Temperatures AND flows (FR-154, RCP Temperature Recorder, and Computer)

3. RCP Number 1 Seal ilP (PI-154A, PI-155A, PI-156A) ACTIONS 1. IF failure of a RCP Number 1 Seal has occurred, THEN refer to AOP-018. DEVICE/SETPOINTS

1. FC-154A, FC-155A, FC-156A / 5 gpm POSSIBLE PLANT EFFECTS 1. Loss of RCS inventory REFERENCES

1. ITS LCO 3.4.4, LCO 3.4.5, LCO 3.4.6 and LCO 3.4.13 2. AOP-018, Reactor Coolant Pump Abnormal Condition

3. CWO B-190628, Sheet 477, Cables K, L, and M I APP-001 Rev. 41 Page 29 of 541 ALARM Rep STANDPIPE HI/LO LVL AUTOMATIC ACTIONS 1. None Applicable CAUSE High *** WILL REFLASH *** 1. Failure of Reactor Coolant Pump Number 1 Seal (after isolation)

2. Failure of Reactor Coolant Pump Number 2 Seal Low 1. Failure of Reactor Coolant Pump Number 3 Seal OBSERVATIONS

1. 2 x 2 Status Light Panel to determine affected Pump 2. Labyrinth Seal ilP (PI-125A, PI-128A, PI-131A) APP-001-C5

3. RCP Seal Leakoff Temperatures AND flows (FR-154, RCP Temperature Recorder, and Computer)

4. RCP Number 1 Seal ilP (PI-154A, PI-155A, PI-156A) ACTIONS 1. IF level is Low, THEN fill the Standpipe using OP-1 01. 2. IF level is High, THEN check for rising level in Reactor Coolant Drain Tank. 3. IF seal failure is indicated, THEN refer to AOP-018. DEVICE/SETPOINTS 1 . LC-406A, LC-407 A, LC-408A 11ft. above normal 2. LC-406B, LC-407B, LC-408B 11ft. below normal POSSIBLE PLANT EFFECTS 1. Loss of RCS inventory REFERENCES 1 . ITS LCO 3.4.4, LCO 3.4.5, LCO 3.4.6 and LCO 3.4.13 2. AOP-018, Reactor Coolant Pump Abnormal Conditions

3. CWD B-190628, Sh 104, Cable J 4. OP-1 01, Reactor Coolant System and Reactor Coolant Pump Startup and Operation I APP-001 Rev. 41 Page 24 of 541 Rev. 17 AOP-018 REACTOR COOLANT PUMP ABNORMAL CONDITIONS Page 5 of INSTRUCTIONS RESPONSE NOT OBTAINED SECTION A REACTOR COOLANT PUMP SEAL FAILURE (Page 1 of 11)

• 1. Check Any RCP #1 Seal Leakoff Flow -GREATER THAN 5.7 GPM 2. Check Either Of The Following Conditions Exist:
• RCP #1 Seal Leakoff Flow On Unaffected RCP(s) -DECREASED IF seal leakoff exceeds 5.7 gpm. THEN Go To Step 2. Go To Step 8. Perform the following:

a. Perform cross-check of all RCP parameters to determine cause of indicated high leakoff flow. b. Observe The NOTE Prior To 41

• RCP Thermal Barrier On Affected RCP(s) -DECREASED Step 1 and Go To The Main Body. Step 1 Of This Procedure

CAUTION To prevent damage to the RCP Seal Stack. the affected RCP Seal Leakoff Isolation valve must be closed between 3 minutes and 5 minutes of stopping the RCP. **************************************************************************

3. Check Plant Status -MODE 1 4. Perform The Following:

a. Trip the reactor b. Trip the affected RCP(s) c. Go To Path-1 while continuing with this procedure.

Stop the affected RCP(s) Observe the CAUTION prior to Step 5 and Go To Step 5. Rev. 17 AOP-018 REACTOR COOLANT PUMP ABNORMAL CONDITIONS

• 16. 17. Page 10 of 41 INSTRUCTIONS RESPONSE NOT OBTAINED SECTION A REACTOR COOLANT PUMP SEAL FAILURE (Page 6 of 11) ... ,,' ..... ..... " Check Calculated

1. 2 Seal Leakoff Flow -LESS THAN l:*I{jppr--'*-

/ .,,,", //.perform the following: \ // a. Place the Plant in Mode 3 f within 8 hours using GP-006. ,/ Normal Plant Shutdown From ) :.::::( Power Operation To Hot ,J I Shutdown. ( / i b. WHEN the plant has been / " placed in Mode 3. THEN stop/ \. the affected RCP (s) . // Check Calculated

1. 2 Seal Leakoff Flow -GREATER THAN 0.5 GPM Go To Step 19. *** ___ o<-...

/ 18. Perform The Following For The Affected RCP: a. Closely monitor RCP Seal parameters

b. Notify Engineering of RCP Seal condition AND instruct them to contact Westinghouse for further instructions.

c. Maintain Seal Injection flow between 8 gpm and 13 gpm d. Go To Step 35

1. 2 SEAL LEAK OFF TO RCOT SEAL LEAK OFF TOVCT #1 SEAL LEAK OFF FROM OTHER RCP'S SEAL INJECTION FROM CHARGING #1 Rep SEAL FAILURE Res-FIGURE-22 1:

i.J 15 IMPELLER INFORMATION USE ONLY #1 SEAL FAILURE As the #1 seal opens up f10wwill increase past the seal and out the #1 leak off line. RCS flow and seal injection flow will turn and flow up the stack giving a thermal barrier low Delta-P (neg.) On a large failure the high flow of hot RCS past the thermal barrier will cause the CCW to flash into steam. This will give high flow spikes on the CCW return from the RCP which in turn will cause FCV-626 to shut. Incrased flow in the failed pump leak off line will pressurize the common return header to the VCT and cause the seal leak off f1owfrom the other pumps to decrease. .. , FCV-626 CCWRETURN FROM THERMAL BARRIER RCSF22 STAND PIPE '12 SEAL LEAK OFF TO RCDT SEAL LEAK OFF TOVCT #1 SEAL LEAK OFF I ...

• FROM OTHER RCP'S SEAL INJECTION FROM CHARGING , CVC-303 CLOSED RCS-FIGURE-23

1. 3 SEAL LEAK OFF , + ' TO SUMP IMPELLER INFORMATION USE ONLY CVC*303 VALVE SHUT When the 303 Valve is shut the major leak path is isolated.

This will relieve the excess pressure on the return header and leak off flow from the other RCP's should return to normal. Additional pressure is placed on the #2 seal causing it to tilt slightly and become a film riding seal similar to the #1 seal. This returns the RCP to near normal values with 5 GPM flowing up past the #2 seal, and a small amount going to the #3 seal. At this point CCW flow through the ther!1lal barrier may be restored by opening FCV*626. The increased flow past the #2 seal will give a high stand pipe level. The net difference will be seal leak off flow going to the RCDT through the stand pipe. (This is treated as RCS leakage). Note: FCV*626 may required several cycles to flush the steam from the lines before it will remain open. FCV*626 RCSF23 'lIIo....-...... CCW Return From Thermal Barrier MAJOR DIFFERENCES RCS-FIGURE-25 THERMAL BARRIER FAILURE SEAL LEAKOFF ill FLOW ALARM SEAL LEAKOFF HI TEMP ALARM THERMAL BARRIER DELTAP LO ALARM X #1 SEAL DELTA P LO ALARM SEAL LEAKOFF LO FLOW ALARM (UNAFFECTED RCPs) THERMAL BARIER COOLING WATER ill X TEMPALARM RCP BEARING TEMP ill ALARM X CCW SURGE TANK HI LEVEL ALARM X R-17 INCREASE OR ALARM X SMALL #1 SEAL FAILURE X X X X INFORMATION USE ONLY LARGE #1 SEAL FAILURE

• X X X X X X X *PRIOR TO SHUTTING CVC-303

1. 015 AA2.01 0011//1//1/

Given the following conditions: QUESTIONS REPORT for AUDIT (Joo2)

• The plant is operating at 8% RTP making preparations for synchronizing to the grid.
• At 0800, Annunciator APP-001-D2, RCP #1 SEAL LEAKOFF HI FLOW, alarms.
• At 0802, the RO reports #1 sealleakoff is 5.2 GPM for RCP 'A'.
• At 0805, the RO reports #1 sealleakoff is now at 5.7 GPM and increasing for RCP 'A', and #1 sealleakoff for RCP 'B' and RCP 'C' have decreased from 2.4 GPM to 1.1 GPM. Which ONE (1) of the following describes the RCP failure and the action required?

A"I #1 seal has failed; Trip the reactor, trip RCP 'A'; enter PATH-1. B. #2 seal has failed; Trip the reactor, trip RCP 'A'; enter PATH-1. C. #1 seal has failed; Trip RCP 'A' and commence plant shutdown lAW GP-006. D. #2 seal has failed; Trip RCP 'A' and commence plant shutdown lAW GP-006. A. Correct. Indications for #1 seal failure and we are in MODE 1. SRO Question 76 Tier 1 Group 1 KIA Importance Rating -SRO 3.5 Ability to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): Cause of RCP failure. Reference(s) -AOP-018, Section A Proposed References to be provided to applicants during examination -None Learning Objective -Question Source -Modified Bank Question History -2002 NRC Exam Question Cognitive Level -Comprehension 10 CFR Part 55 Content -43 Comments -Category 1: Category 3: Category 5: Category 7: Monday, June 09, 2008 11 :03:59 AM Category 2: Category 4: Category 6: Category 8: 1 HLC-08 NRC Written Exam 4. Given the following: -The plant is operating at 100% RTP. -RCS boron concentration is 600 ppm. -Earlier in the shift, VCT level transmitter L T -115 failed LOW. -All actions of AOP-003, MALFUNCTION OF REACTOR MAKEUP CONTROL, have been completed. -VCT level is currently 30 inches as indicated on level transmitter L T -112. -I&C reports they have deenergized L T -112 inadvertently. Which ONE (1) of the following describes how the actual VCT level and T AVG would respond? (Assume no operator action) A. VCT level increases; T AVG increases. B. VCT level decreases; T AVG increases. C. VCT level increases; T AVG decreases. D. VCT level decreases; T AVG decreases. V DT ltwl ote.U'*.Q56 \.50 nO+-,?\w.<.s'l'v:M. No OV\. \;"c0 C\..v.... CC\;.A-\--\-0 "VC*T 90\"'::' 1 tIL v(h",6 '30 c10uAV\. 4 I'A-V{'- \> '(\(7\- Ie, £vx.1f-. \J L. 'T Ov-C!o 1"",,\4>.. ) Ii\--i 6- \I) VI c.. 'l'tcx.5f. HLC-08 NRC Written Exam 4. 000022 AAl.Ol OOllLOSS OF RX COOL MAKE/1I1I3.4/3.3IROIHIGHJNIAIFARLEY -2001lCVCS-004 Given the following: -The plant is operating at 100% RTP. -RCS boron concentration is 600 ppm. -Earlier in the shift, VCT level transmitter L T -115 failed LOW. -All actions of AOP-003, MALFUNCTION OF REACTOR MAKEUP CONTROL, have been completed. -VCT level is currently 30 inches as indicated on level transmitter L T -112. -I&C reports they have deenergized LT-112 inadvertently. Which ONE (1) of the following describes how the actual VCT level and T AVG would respond? (Assume no operator action) A. VCT level increases; T AVG increases. B. VCT level decreases; T AVG increases. C:-" VCT level increases; T AVG decreases. D. VCT level decreases; T AVG decreases. The correct answer is C. A: Incorrect -RWST will be supplying the charging pump suction with water of 2000+ ppm boron, this will cause T AVG to decrease. B: Incorrect -T AVG will NOT increase due to boration from the RWST. C: Correct -Letdown will still be going to the VCT causing level to increase; boration from the RWST will be causing T AVG to decrease. Having both level transmitters low will open LCV-115B and shut LCV-115C. This swaps charging pump suction from the VCT to the RWST. 0: Incorrect -VCT level will NOT decrease, there is no outlet flow from the VCT due to LCV-115C going shut and letdown still going to the VCT. Tuesday, June 17, 2008 1 :21 :00 PM 4 HLC-08 NRC Written Exam Exam Question Number: 4

Reference:

SO-021, eves, Pages 41, 56, Figures 1, 17a, 6; OWP-005, CVCS, Page 53. KA Statement: Ability to operate and / or monitor the following as they apply to the Loss of Reactor Coolant Makeup: eves letdown and charging. History: Modified from Farley 2001 NRC exam, Changed distractors/answer to T AVG VS Rx power. Tuesday, June 17,2008 1 :21 :00 PM 5 SD-021 CHEMICAL AND VOLUME CONTROL SYSTEM be accommodated in the VCT. LCV-115A is a three-way valve which is controlled by a AUTO/MANUAL proportional controller which uses VCT level (normally LT-112) as an input when selected to AUTO. If level in the VCT reaches 51.6" the valve will be removed from its controller and fully diverted to the CVCS holdup tanks. The valve can also be overridden with the switch on RTGB. Upon a loss of air, LCV-lISA will fail to the VCT position. If LCV-115A is aligned to divert letdown from the VCT, flow is directed to one of three CVCS Holdup Tanks. Alarms with the CVCS Holdup Tanks annunciate at the Waste Disposal Boron Recycle Panel. The CVCS Holdup Tanks and their instrumentation are discussed in more detail in the Waste Disposal System Description. 5.2.3 Emergency Makeup to Charging Suction (LCV-115B) -LCV -115B is an air-operated valve which is controlled by a three-position AUTO-OPEN) switch on the RTGB. In AUTO, this valve will automatically open to provide emergency makeup from the RWST to the VCT if level drops below 12.4". LCV -115B has a backup air supply from the compressed air bottles in the charging pump room. There is also a three-position switch located on the Charging Pump Panel in the Charging Pump Room. The three positions are CLOSED-AUTO-OPEN. This key operated switch will always be maintained in the AUTO position. While in AUTO this control scheme (totally independent of the normal interlock with LCV -115C/auto swap over) will open LCV-115 if both LCV-115C and LCV-115B are closed at the same time for 60 seconds. An AC powered solenoid was added to LCV -115B and it is powered by the DS Bus. This new control scheme was added to ensure the running charging pump will have a suction source during a smart fire event. 5.2.4 VCT Outlet Valve (LCV-115C) .;;11' LCV-115C is a motor operated valve, controlled from the RTGB by a three position (CLOSE, AUTO, OPEN) switch. In AUTO, its position is controlled by LT-112 and LT-115 for the emergency swapover to the RWST via LCV-115B. The power supply to LCV-115C is MCC-6. 5.2.5 Makeup System Components eves Boric Acid Flow Control Valve, FCV-1I3A, BA TO BLENDER, is controlled by a three position (OPEN, AUTO, CLOSE) RTGB switch. In AUTO, the position is controlled by AUTO/MANUAL RTGB controller FCV-113A, BORIC ACID FLOW. Operation with the potentiometer set for greater than 9.0 should be avoided since this could result in boric acid flowrates of greater than 10.0 gpm which are beyond the indicating range of FR-113 and YIC-113. This valve is used for boration and normal makeup. It regulates the boric acid flow to the blender and fails open. Page 41 of71 Revision 1 0 INFORMATION USE ONLY SD-021 CHEMICAL AND VOLUME CONTROL SYSTEM LT-115 L T -115 Fails L T -112 Fails L T -112-Fails Fails LOW IDGH LOW IDGH ERFIS LT-1I2 Decreasing BAD; <0 BAD; >100 ERFIS LT-1I5 BAD; <0 BAD; >60 No effect RTGB VCT LvI <0 >60 No effect LCV-1I5A Closed Open Closed Open LC-115A No Demand No Demand No Demand Full Demand Auto Makeup Start; Not Available Still available Still Available; Continuous cycling Auto Emergency Still possible Not possible Still possible Not possible . Switchover APP-003-E3 Received Received No No Overall Effect Continuous VCT level Hardest to VCT level will auto makeup decreases due recognize decrease due as long as to LCV-115A since no real to LCV-1I5A system in open off of change to diverting. An AUTO. LT-1I5 Hi system other auto makeup VCT fills level backup. than LCV-will occur and and LCV-Since no auto 1I5A will not restore VCT 115A diverts emergency divert until level. The off ofLT-switchover is the LT-115 makeup will 112 control. possible, if Hi level take longer operator backup is than normal. action is not reached. Will Once makeup taken will only see if stops, VCT lose charging monitor level will pump suction. ERFIS. decrease and cause a repeated cycling of the auto makeup. 6.5 Emergency Operation eves The CVCS is used in the Dedicated Shutdown procedures and the Emergency Operating procedures to control RCS inventory and maintain seal injection. Its use can be from inside the control room or outside the control room in either set of procedures. For example, a fire in the charging pump room would render normal charging useless, but the Dedicated Shutdown procedures provide a means to shutdown and cooldown the Page 56 of7! Revision 1 0 INFORMATION USE ONLY TOPRT .... .....---. LOOPt LOOPt HOT LEG AUX SPRA Y4V1' .14 I LOOP 2 COLD LEG From Loop 2 Cold Leg eves FLOW DIAGRAM CVCS-FIGURE-l From RCP Seals Charging Pulsation Pumps Dampener 286 '\ To Rep Seals 1 INFORMATION USE ONLY From RWST 245 ... Mixed Bed I u U Demins ::orating 225 Demins .. 216 237A Cation J I Bed Demin. 352 353 FromBA Transfer Pumps

\2/K VCT LEVEL CONTROL CVCS-FIGURE-17A FROM eve DEMINS OVERRIDES LCll2AAT 51.6 AND RETURNS CONTROL AT 45.6" ,----------------------, " NORMAL

• J LT11mT115 1 AUTO U LCV-115C INTERLOCKED SO BOTH CAN'T CLOSED IN AUTOMATIC FROM : SWITCH LOCATED IN : t--II .....

I RACK 19 I MAKEUP I . FC -114B ) .., I -8 I CONTROLLER

LC115A COMPAR* I ATOR I I LC115C HIALARM48.6

__ '_. DUAL I COMPAR'f-__ ALARM 17.2" t r I LCl15B ---4 AUTO MAKEUP AT 1.... DUAL 20.2" AND STOP 'r AT 24.4" I I B-------." I COMPARATOR" ,,) , , " " !---... I-FROM AUTO MAKEUP LO-LO (12.4") CLOSES LCV-115C AND OPENS LCV-115B IN AUTO LCV-115B FROM RWST

• IF SWITCH IN LT-112; LT-112 CONTROLS ALL COMPARATORS

& CONTROLLERS. IF SWITCH IN LT-115: LT-115 CONTROLS ALL COMPARATORS & CONTROLLERS CHARGING PUMP SUCTION /cvcsf11a / INFORMATION USE ONLY AUTOMATIC (NORMAL) MAKEUP FLOWPATH cvcs-FIGURE-6 FROM lCV-115A CHARGING PUMPS 356 LCV-115B CVCS MAKEUP CONTROL ... ---., --., I I I I I l--..:......ji ..... Y-FCV-113A FROM RWST B 269 r 358 HEAT TRACING INFORMATION USE ONLY FROM PRIMARY WATER PUMPS FROM BORIC ..... ... ACID TRANSFER PUMPS REFERENCE USE OWP Title: CVC-18 Page _1_ of _2_ LT-115, Volume Control Tank Level{ TC "LT-llS f Volume Control Tank Level" \f C \1 "1" } 1. This revision has been verified to be the latest revision available.

2. 3. 4. 5. 6. Initial Signature Date Name (Print) System: CVC W/O No:, _________

_ Component: L T -115, Volume Control Tank Level Scope of Work: Calibrate/repair L T -115. Testing required on redundant equipment prior to rendering component inoperable: N/A Precaution: _---1) When placing L T -112 in operation, the selector switch should be operated swiftly to prevent prolonged loss of level signal from both loops. Loss of both ? 7. 8. signals could cause LCV-115B to open r--=--2) When selected to the L T -112 pOSition, the RTGB and ERFIS indications will be driven by LT-112. 3) If mechanical work will not be performed, only the electrical lineup is required.

4) If mechanical work will be performed, both the electrical and the mechanical lineups are required.

5) Complete the electrical lineup for maintenance prior to performing the mechanical lineup. 6) This activity has been screened in accordance with PLP-037 criteria and determined to be outside the bounds of an Infrequently Performed Test or Evolution.

Valve/Breaker/Switch lineup has been completed. Clearance Issued (If applicable) / Signature Date Clearance No: _________ _ 9. Testing required on redundant equipment while component is inoperable.

10. 11. 12. N/A I&C Maintenance lineup completed.

Clearance removed and Valve/Breaker/Switch lineup restored to normal. Post Maintenance Testing. 1) Verify calibration of component repaired or replaced in accordance with PIC-002. 1 OWP-005 (c vcs) Rev. 57 N/A / N/A Signature Date / Signature Date / Signature Date Page 53 of 961 \ ; , i \ \ A -Correct, need a ctmt pressure of 27 psig to start CS pumps with an LOSP. If all systems function normally, an SI signal is recieved at 4 psig. B -Incorrect, Requires 27 psig on 2/4 detectors. C -Incorrect, Must also have an SI signal present. D -Incorrect, An SI signal alone will cause the Containment Fans Coolers to operate in slow speed but will not start a spray pump. Source: New Answer: A --... -... ----3 . . 1 FI1;eu= 2. /{/,-(( c 01 Given the following conditions on Unit 2: -Reactor is at 100 % steady-state power. -RCS boron concentration is 600 ppm. -VCT level is currently 40% as indicated on level transmitter LT -112. -Earlier in the shift VCT level transmitter LT-115 failed low and has been released to I&C for repair. -Make up Mode Control Switch is in STOP -The I&C person has just left the control room to electric all y isolate LT -115. I&C reports they have deenergized LT-112 inadvertently. Which ONE of the following describes how the VCT level and reactor power would respond? (Assume no operator action) A. VCT level increases; reactor power remains the same. B. VCT level decreases; reactor power increases. C. VCT level increases; reactor power decreases. D. VCT level decreases; reactor power decreases. A -Incorrect, RWST will be supplying the charging pump suction with water with 2000+ ppm boron, this will cause reactor power to decrease. B -Incorrect, Reactor power will not increase due to boration from the RWST. C -Correct, Letdown will still be going to the VCT causing level to increase; boration from the RWST will be causing power to decrease. Having both level transmitters low will open LCV-1I5B & D and shut LCV-1I5C & E this swaps charging pump suction from the VCT to the RWST. D -Incorrect, VCT level will not decrease, there is not out let flow from the VCT due to LCV-115C and LCV-115E going shut and letdown still going to the VCT. Source: Modified from Vogtle NRC Exam 2001 HLC-08 NRC Written Exam 5. Given the following: -Following a LOCA, the plant is in Cold Leg Recirculation lAW EPP-9, TRANSFER TO COLD LEG RECIRCULATION. -RHR Pump "A" is in service providing core cooling. -RHR Pump discharge pressure is fluctuating by 40 PSIG and flow is approximately 3350 GPM and oscillating. -APP-001-A7, RHR HX LO FLOW, has alarmed and cleared several times. -APP-003-D3, PRT HIILO LVL alarm has just illuminated. Which single malfunction could be causing the event? A. ECCS Sump screens are clogging. B. RHR-752A, RHR PUMP A SUCTION VALVE has drifted partially shut. C. RHR-706, RHR RELIEF VALVE is cycling. D. SI-887, RHR RECIRCULATION TO RWST VALVE disc has separated from its valve stem. \'{-.'!Le 'c." o\'\Jtl!1Q5 ()vv\ :;'clAA-"-, r ct.) nor 2)ee... 'i\-eVAy < .. xe&}!} '" d(.'.>'\-fCL(,/..-O. 5 HLC-08 NRC Written Exam 5. 000025 AA2.06 OOllLOSS OF RHR/l/l/3.2/3.4IRO/HIGH/N/A/NEW -200S/AOP-020-004 Given the following: -Following a LOCA, the plant is in Cold Leg Recirculation lAW EPP-9, TRANSFER TO COLD LEG RECIRCULATION. -RHR Pump "A" is in service providing core cooling. -RHR Pump discharge pressure is fluctuating by 40 PSIG and flow is approximately 3350 GPM and oscillating. -APP-001-A7, RHR HX LO FLOW, has alarmed and cleared several times. -APP-003-D3, PRT HIILO LVL alarm has just illuminated. Which single malfunction could be causing the event? A. ECCS Sump screens are clogging. B. RHR-752A, RHR PUMP A SUCTION VALVE has drifted partially shut. RHR-706, RHR RELIEF VALVE is cycling. D. SI-887, RHR RECIRCULATION TO RWST VALVE disc has separated from its valve stem. The correct answer is C. A: Incorrect -Sump screen clogging would cause RHR flow to decrease NOT fluctuate. PRT level indicators make this condition NOT the cause. B: Incorrect -RHR-752A is closed in Cold Leg Recirculation. C: Correct -Normal flow rate (NOT a suction problem), with pressure oscillations and PRT level indications. D: Incorrect -SI-887 is isolated from RHR during Cold Leg Recirc. (SI-863A and 863B are closed) Exam Question Number: 5

Reference:

APP-001-A7; APP-003-D3; SD-003, RHR, Figure 3. KA Statement: Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System: Existence of proper RHR overpressure protection. History: New -Written for HLC-08 NRC exam. Tuesday, June 17, 20081:21:01 PM 6 ALARM RHR HX LO FLOW AUTOMATIC ACTIONS 1. None Applicable CAUSE 1. Loss of RHR Pump 2. Loss of suction head (Le., Loss of water from RCS) 3. Closure of system flowpath valves 4. Intentional operator action OBSERVATIONS

1. RHR Flow (FI-605) 2. RHR Pump status 2. Position of HCV-758 4. Position of FCV-605 5. Position of system flowpath valves 6. Pressurizer Level (LI-459A, LI-460, LI-461 , and LI-462) 7. RCS Loop Standpipe Levels (LI-403, LI-404) ACTIONS 1. IF alarm due to intentional operator action, THEN no further action is required.

APP-001-A7

2. IF alarm due to flow control valve operation, THEN verify flow set between 2800 gpm and 3750 gpm. 3. IF flow can NOT be restored to greater than 2800 gpm via flow control valve change, THEN refer to AOP-020. OEVICE/SETPOINTS 1 . FC-605A I 3000 gpm POSSIBLE PLANT EFFECTS 1. Loss of decay heat removal REFERENCES

1. ITS LCO 3.4.6, LCO 3.4.7, LCO 3.4.8, LCO 3.9.4 and LCO 3.9.5 2. AOP-020, Loss of Residual Heat Removal System (Shutdown Cooling) 3. CWO B-190628, Sheet 487, Cable S I APP-001 Rev. 41 Page 11 of 541 APP-003-D3 ALARM PRT HI/LO LVL *** WILL REFLASH *** AUTOMATIC ACTIONS 1. Not Applicable CAUSE High 1. Excessive makeup water added 2. In leakage from Makeup Water, Pressurizer Relief Valves, Pressurizer Safety Valves, RHR Loop Relief Valves, Letdown Relief Valves, Seal Water Return Relief Valve, SI Test Line Relief Valve, or SI Cold Leg Injection Header Relief Valve 3. Opening of Pressurizer Safety or PORV Low 1. Leakage from PRT to the Reactor Coolant Drain Tank or other area. 2. Excessive draining.

OBSERVATIONS

1. PRT Level (L1-470), Pressure (PI-472), and Temperature (TI-471) 2. Pressurizer Safety Valve Line Temperatures (TI-465, TI-467, TI-469) 3. PORV Discharge Line Temperature (TI-463) ACTIONS 1. IF a PZR PORV or Safety fails open while greater than 350°F, THEN Refer To Path-1. 2. IF level is high, THEN drain the PRT using OP-103. 3. IF level is low, THEN add Primary Water to the PRT using OP-103. DEVICE/SETPOINTS

1. LC-470 I 83% 2. LC-470 I 68% POSSIBLE PLANT EFFECTS 1. None Applicable REFERENCES 1 . Path-1, EOP Network 2. OP-103, Pressurizer Relief Tank Control System 3. CWD B-190628, Sheet 461, Cable M, N I APP-003 Rev. 37 Page 33 of 531 COLD LEG RECIRC -RHR FLOW> 1200 GPM, RCS<125 PSIG RHR-FIGURE-3

,------,-------------------------11/1-1

---

.., TOSI PUMP B&C

• SUCTIONS I SI*891 C SI*891 D RHR*764 r-evcs I' ___ .J l+ RHR* FCV*605 TO PRESSURIZER RHR*757A <;: .... o ;;-!i! n RHR*754B MINI FLOW RECIRC RELIEF TANK RHR*743 RHR*744B TO RC lOOP 1 SI.:a7sA COLD lEG SI*876B SI*875B TO RC LOOP 3 COLD lEG TO RC lOOP 2 COLD lEG SI*887 RHR*752B RHR PUMP B RHR PUMP A -

-FROM CONTAINMENT SUMP RHR HEAT-UP LINE INFORMATION USE ONLY TO SIAND CONTAINMENT SPRAY PUMP SUCTIONS .'4 .'4 . FROM RWST .... Ul o FROM RC LOOP 2 HOT lEG HLC-08 NRC Written Exam 6. Given the following: -The Crew is responding to a LBLOCA lAW PATH-1 and are unable to start any CCW pumps. -AOP-014, CCW SYSTEM MALFUNCTION, Section "C", CCW PUMP DISCHARGE PRESSURE LOW is being performed. -An AO was dispatched to perform Attachment 1, EMERGENCY COOLING TO CHARGING PUMPS. -Charging Pump "B" is running with speed controller in MANUAL set at 75%. -All RCP Thermal Barrier DP indications are at approximately 3.5 inches. -Step 9 of AOP-014, Section "c" directs raising Charging Pump speed to obtain RCP Thermal Barrier DP of > 5 inches. Which ONE (1) of the following is the reason for raising the RCP Thermal Barrier DP to > 5 inches? A. The increased charging rate ensures RCP seal leakage will be minimized. B. ONLY seal injection is available for cooling, since there is no CCW to the Thermal Barriers. C. The increased charging rate ensures the Charging Pump temperatures remain low until Attachment 1 is completed. D. Increasing the Thermal Barrier DP is in preparation for restarting the RCP(s) after CCW is restored. \"D\\ bs CvJ'.'( sc:... G:vc-r \o-vcc\,\L U:1l-fJ-, 6 HLC-08 NRC Written Exam 6. 000026 AAl.02 OOllLOSS OF CCW/l/l/3.2/3.3IROILOW/N/A/NEW -2008/AOP-014-003 Given the following: -The Crew is responding to a LBLOCA lAW PATH-1 and are unable to start any CCW pumps. -AOP-014, CCW SYSTEM MALFUNCTION, Section "C", CCW PUMP DISCHARGE PRESSURE LOW is being performed. -An AO was dispatched to perform Attachment 1, EMERGENCY COOLING TO CHARGING PUMPS. -Charging Pump "B" is running with speed controller in MANUAL set at 75%. -All RCP Thermal Barrier DP indications are at approximately 3.5 inches. -Step 9 of AOP-014, Section "C" directs raising Charging Pump speed to obtain RCP Thermal Barrier DP of > 5 inches. Which ONE (1) of the following is the reason for raising the RCP Thermal Barrier DP to > 5 inches? A. The increased charging rate ensures RCP seal leakage will be minimized. B:' ONLY seal injection is available for cooling, since there is no CCW to the Thermal Barriers. C. The increased charging rate ensures the Charging Pump temperatures remain low until Attachment 1 is completed. D. Increasing the Thermal Barrier DP is in preparation for restarting the RCP(s) after CCW is restored. The correct answer is B. A: Incorrect -RCP Seal leakage is NOT a concern for LBLOCA, but is a step for increased seal leakage. B: Correct -Thermal Barrier Delta P is the only means of ensuring pump seals are receiving adequate cooling. C: Incorrect -Step 9 RNO has direction to wait until Attachment 1 is complete before reducing Charging Pump speed to minimum, but NOT for Charging Pump temperature. D: Incorrect -Raising Thermal Barrier Delta P to minimum is a step to start RCPs, but no restart of RCP will occur in this situation. Tuesday, June 17,20081 :21 :01 PM 7 HLC-08 NRC Written Exam Exam Question Number: 6

Reference:

AOP-014, Section C, Page 40; AOP-014 SD, Page 24. KA Statement: Ability to operate and / or monitor the following as they apply to the Loss of Component Cooling Water: Loads on the CCWS in the control room. History: New -Written for HLC-08 NRC exam. Tuesday, June 17, 20081 :21 :01 PM 8 Rev. 24 AOP-014 COMPONENT COOLING WATER SYSTEM MALFUNCTION Page 40 of 104 INSTRUCTIONS RESPONSE NOT OBTAINED SECTION C CCW PUMP DISCHARGE PRESSURE LOW (Page S of 7) Check RCP Thermal Barrier -GREATER THAN S INCHES

• PI-13lA
• PI-128A
• PI-l2SA 10. Notify Chemistry Personnel To Stop Any Sampling In Progress *11. Determine If Emergency Cooling To Spent Fuel Pit Heat Exchanger Is Required As Follows: a. Check APP-036-B4.

SPENT FUEL PIT HI TEMP -ILLUMINATED

b. Align emergency cooling to the Spent Fuel Pit Heat Exchanger using Attachment

2. Emergency Cooling To Spent Fuel Pit Heat Exchanger.

while continuing with this procedure Increase Charging Pump speed to maintain RCP Thermal Barrier greater than S inches. WHEN Attachment 1 has been completed. THEN perform the following:

a. Reduce the running Charging Pump to minimum speed. b. Locally throttle RCP SEAL WATER FLOW CONTROL VALVE(s) to assure at least 6 gpm flow to each RCP.

• CVC-297 A
• CVC-297B
• CVC-297C a. IF at any time APP-036-B4 illuminates.

THEN perform Step 11. b. Go To Step 12. 10 11 12 13 14 15 16 BASIS DOCUMENT, COMPONENT COOLING WATER SYSTEM MALFUNCTION Description While starting and stopping Charging Pumps, the operator should observe Seal Injection flow by checking Thermal Barrier indications. During this period no CCW cooling to the Thermal Barrier exists, therefore the only seal cooling is provided by intermittent seal injection. (CR 95-1752, see Section A, step 34 for additional information) Chemistry is notified to stop sampling due to the lack of COOling to the heat exchangers. This steps determines if cooling must be supplied to the SFP Heat Exchanger. The step checks temperature above the high temperature alarm setpoint, then commences efforts to establish cooling. It is not expected that the SFP will reach elevated temperatures for an extended period of time. This procedure assumes that refueling efforts are NOT in progress. Should those efforts be underway, plant procedures are in place to more rapidly establish emergency cooling in the event that normal cooling is lost. This step acts as a hold point until a CCW Pump is started. At this point all actions to restart a pump have been directed and emergency cooling is established where possible. If Charging Pumps are being supplied by emergency cooling they may now be restored to normal cooling. Engineering is consulted at this point to determine what restoration should be used based on the type of cooling used. The pump coolers will have been contaminated with high chloride water and may require extensive flushing. This step determines if Charging and letdown should be reestablished. If the RCS is above 150°F charging will be required. If RHR is not in the Core Cooling Mode letdown will be required. If these are required subsequent steps will establish. If not, the operator is transitioned around the steps. If charging is required, pumps are started (see above). If letdown is required it is placed in service using the appropriate procedure. This step is entered if previous steps determined that charging and letdown must be placed in service. I AOP-014-BD Rev. 24 Page 24 of 341 HLC-08 NRC Written Exam 7. Given the following: -The Reactor failed to automatically trip when the Reactor Coolant Pumps tripped. -APP-005-F6, AMSAC TRI P alarm is illuminated. Which ONE (1) of the following will result? A. Turbine trip and and ALL AFW Pumps start. B. Reactor trip and and ALL AFW Pumps start. C. Reactor trip and and ONLY MDAFW Pumps start. D. Turbine trip and and ONLY MDAFW Pumps start. LtfY)cVtl W\5 L<;:I!l?lA..\t\ C CVU $*.. eM \y e.., tv\S) A Fw '\> tt-w.{> 5> 'b 7 HLC-08 NRC Written Exam 7.00002902.1.28 001/ATWS/1/1I4.1I4.1IRO/HI0H/NINNEW -2008/AMSAC-006 Given the following: -The Reactor failed to automatically trip when the Reactor Coolant Pumps tripped. -APP-005-F6, AM8AC TRIP alarm is illuminated. Which ONE (1) of the following will result? A'I Turbine trip and and ALL AFW Pumps start. B. Reactor trip and and ALL AFW Pumps start. C. Reactor trip and and ONLY MOAFW Pumps start. D. Turbine trip and and ONLY MOAFW Pumps start. The correct answer is A. A: Correct -AMSAC trips the Turbine and starts BOTH MDAFW and the SDAFW Pump. B: Incorrect -AMSAC does NOT trip the Reactor at RNP. C: Incorrect -AMSAC starts ALL AFW Pumps, NOT just the MDAFW Pumps. D: Incorrect -AMSAC does NOT trip the Reactor at RNP and the circuit starts ALL AFW Pumps .. Exam Question Number: 7

Reference:

APP-005-F6; 80-062, AM8AC, Figure 1. KA Statement: Knowledge of the purpose and function of major system components and controls. History: New -Written for HLC-08 NRC exam. Wednesday, June 18, 200810:14:50 AM 9 ALARM AMSACTRIP AUTOMATIC ACTIONS 1. Motor and Steam Driven AFW Pumps Start. 2. Turbine Trips. CAUSE APP-005-F6

1. Low-Low Level in two out of three Steam Generators when above 40% turbine load. OBSERVATIONS

1. S/G "A" Level (LI-474) 2. S/G "B" Level (LI-485) 3. S/G "C" Level (LI-496) 4. Turbine First Stage Press (PI-446 & PI-447) ACTIONS 1. IF a reactor trip has occurred, THEN refer to EOP Network. DEVICE/SETPOINTS

1. AMSAC PROCESSOR

-S/G Level Narrow Range L T -47 4, L T -485, L T -496/11 % -Turbine First Stage Press PT-446, PT-447 1 191 psig (35%) POSSIBLE PLANT EFFECTS 1. Plant Shutdown REFERENCES

1. EOP Network 2. CWD B-190628, Sheet 1731 I APP-005 Rev. 29 Page 40 of 40 I LT-474 LT-485 2/3 <11% AMSAC LOGIC DRAWING AMSAC-FIGURE-l LT-496 PT-446 l i 360 Second T.O. Relay J<J 25 Second T.O. Relay \...t3t...

RIP/AFW START 2/2 35% ARM IX INFORMATION USE ONLY PT-447 RTGB BYPASS amsacf01 HLC-08 NRC Written Exam 8. Given the following: -The crew is responding to a SGTR in S/G "A" with a loss of offsite power. -Supplement G is in progress. -ALL S/G pressures are 1025 PSIG. -ALL S/G PORVs are partially OPEN. -The step in Supplement G which requires Steam Flow to be isolated from the ruptured S/G has just been reached. Which ONE (1) of the following describes the action that is required regarding the ruptured S/G PORV lAW Supplement G? A. Raise the RUPTURED S/G PORV controller's setpoint to MAXIMUM. B. Verify the RUPTURED S/G PORV controller's setpoint at 1035 PSIG. C. Reduce the RUPTURED S/G PORV controller's setpoint to 1000 PSIG. D. Verify the RUPTURED S/G PORV controllers' setpoint is 15 PSIG less than the INTACT S/G PORV controllers. 8 A II \,'1 \\<f-y ? Luheovt wolA\b \'CL/'s.e 1'f\c,)<.I\<v\.u,Vv -: bo y\ t.y eve" do s ? HLC-08 NRC Written Exam 8. 000038 G2.1.20 001lSGTR/1I114.6/4.6IROILOWINIAIRNP BANKIPATH-2-003 Given the following: -The crew is responding to a SGTR in S/G "A" with a loss of offsite power. -Supplement G is in progress. -ALL S/G pressures are 1025 PSIG. -ALL S/G PORVs are partially OPEN. -The step in Supplement G which requires Steam Flow to be isolated from the ruptured S/G has just been reached. Which ONE (1) of the following describes the action that is required regarding the ruptured S/G PORV lAW Supplement G? A. Raise the RUPTURED S/G PORV controller's setpoint to MAXIMUM. 8:1 Verify the RUPTURED S/G PORV controller's setpoint at 1035 PSIG. C. Reduce the RUPTURED S/G PORV controller's setpoint to 1000 PSIG. D. Verify the RUPTURED S/G PORV controllers' setpoint is 15 PSIG less than the INTACT S/G PORV controllers. The correct answer is B. A: Incorrect -Placing the ruptured S/G PORV to MAXIMUM is undesirable as the PORV will NOT OPEN and pressure may reach S/G safety valve setpoint. B: Correct -The setpoint on the ruptured S/G PORV is increased to 1035 PSIG from 1025 PSIG. This minimizes the potential for radiological release and ensures the PORV is maintained available to prevent challenging the S/G safety valves. C: Incorrect -Reducing the RUPTURED S/G PORV's setpoint would result in MORE, NOT less release of radioactive fluid to the atmosphere. D: Incorrect -Reducing the RUPTURED S/G's PORV setpoint to less than the intact S/Gs would result in more steam flow from the contaminated S/G, rather than from the intact S/Gs. Exam Question Number: 8

Reference:

Supplement G, Page 37; Supplement G BD, Pages 4 and 7. KA Statement: Ability to interpret and execute procedure steps. History: Direct from Bank. Wednesday, June 18, 2008 10:14:50 AM 10 EPP-Supplements INSTRUCTIONS SUPPLEMENTS CONTINUOUS USE Supplement G Rev. 35 Page 37 of 89 RESPONSE NOT OBTAINED Steam Generator Isolation (Page 1 of 12) vt. Go To Appropriate Step From Following Table: S/G TO BE ISOLATED STEP S/G A 2 S/G B 18 S/G C 34 2. Check S/G A -FAULTED 3. Verify VI-3A, MSIV -CLOSED WHEN Tavg less than 547°F, THEN perform Steps 3 AND 4. WHEN S/G A level is greater than 8% [18%J, THEN observe the NOTE prior to Step 5 and perform Steps 5 though 9. _.--\ Verify ruptured STEAM LINE PORV",,", setpoint at 1035 psig using ) atus Board ."' . ..------Go To Step 11. 4. Verify MS-353A, MSIV VI-3A BYP -CLOSED Local operation of the FRV and B/P valves below is via reverse acting handwheels.

5. Verify FRV A -CLOSED 6. Verify FRV A BYP -CLOSED

8. 9. 10. 11. 12. 13. 14. Supplement G -This supplement provides the Operator with a listing of valves to be closed to isolate a Steam Generator.

The Supplement differentiates between the two types of SIG problems to ensure that the proper actions are taken to safely isolate the S/G. Therefore the Supplement ensures that the MSIV on a ruptured SIG is not closed prior to RCS temperature being less than 54?DF thereby preventing an inadvertent lifting of the Steam Line PORV while the MSIV on a faulted SIG is immediately closed to minimize the uncontrolled cooldown. The UFSAR credits SIG isolation within 30 minutes for the consequences of the Off-Site Dose release. From the original licensing basis this is considered to occur when the MISV on the affected SIG is closed (Original FSAR chapter 14 description). Supplement H -This supplement provides the Operator with a listing of valves that require local checks for Phase A Verification during a loss of DC. The intent is that the Operator will perform normal verification for those valves that have indication in the Control Room. Then also dispatch an Operator to locally verify the valves listed in the Supplement. The Supplement only checks those valves that do not have redundant isolation with indication in the Control Room. Supplement I -This supplement provides the Operator with an abbreviated lineup for placing the RHR System in the Core Cooling Mode. This Supplement is referenced in numerous procedures throughout the EOP Network during evolutions involving an RCS cooldown. This supplement provides the minimum items necessary to place the RHR System in the Core Cooling Mode in an expeditious manner, reflecting the priorities of the EOP Network. Supplement J -This supplement provides a methodology for determining RHR flow in the event that FT-605 becomes inoperable. FT -605 does not have redundant indication. Supplement K -This supplement provides valve alignments that may be used to enhance Auxiliary Spray flow when required in the EOP Network. Various procedures within the EOPs require the use of Auxiliary Spray. Should the flow rate available from charging be insufficient to reduce pressure, this supplement may be used to increase flow by forcing all charging flow through the Auxiliary Spray valve instead of both Auxiliary Spray and the normal Charging Line. To provide for a finer control of the spray flow into the Pressurizer and the subsequent RCS pressure changes, the Charging Pump Speed can be varied or the normal Pressurizer spray valves from the RCPs can be modulated to provide a recirc path back to the RCS cold leg. This supplement is not required to be used. It is only included should the operator need to reference the steps. Supplement L -This supplement provides the guidance for verifying automatic equipment start following an SI. Its use is directed from within procedures which may require an SI, but from which exiting to Path is not desired. An Operator would use the Supplement to verify equipment starts while the Control Room Team remains in the primary procedure. This occurs primarily in FRPs, such as FRP-S.1 and FRP-H.1. Supplement M -This supplement provides a list of components in the secondary that must be secured following isolation of SW to the Turbine Building. The Supplement ensures that the MSIV on a ruptured SIG is not closed prior to RCS temperature being less than 547°F thereby preventing an inadvertent lifting of the Steam Line PORV. This list is too lengthy to be provided in the Path. Supplement N -This supplement provides the steps necessary to isolate a ruptured SIG from the intact SIGs on failure of the ruptured SIG MSIV to close when demanded. The steps are derived from WOG ERG, E-3. These steps had formerly been in Path-2, however, due to difficulties experienced by some crews in performing the steps, they were moved to this supplement in order to provide more detailed information in the steps than can be provided in the path decision and action blocks. These steps are necessary if an MSIV fails to close since the WOG strategy requires that the ruptured SIG pressure remain elevated in order to limit the amount of depressurization required to stop the primary to secondary leakage. By isolating the ruptured SIG from the intact SIGs, the RCS may be cooled by steaming the intact SIGs to atmosphere without effecting the pressure in the ruptured S/G. I EPP-SUPP-BD Rev. 35 Page 4 of 10 I RNP Basis STEP Supplement G This supplement is transitioned to from various pOints in the EOP Network. It contains instructions to isolate the S/Gs. Placing these instructions in a supplement prevents repeating this information a number of times in the EOP Network. The supplement is structured such that the Operator may move quickly to the appropriate step to isolate the faulted or ruptured S/G. An RNO is provided early in each section to provide for steps that are time or condition dependent when a S/G Tube Rupture has occurred. Generally, the Operator will stay in the left (AER) column for a Faulted S/G and enter the RNO at step a for a Ruptured S/G. The step for verifying the S/G PORV closed is only reached if the S/G is faulted since for a rupture the action is to verify the correct setpoint. During a fault or a ruptured-faulted, the pressure will always be below the setpoint and therefore should be closed. Once dry out occurs for the affected S/G, actions are directed to dump steam from intact generators to control RCS repressurization and temperature increase. The MSIV Above and Below Seat drains are listed as a check step with an RNO to direct the "Dedicated Operator" to close the appropriate valve. The intent of these steps is to ensure that the valves are closed by the most expeditious means. It is expected that the CRSS will know if any have been opened since these valves are normally locked closed and are controlled under the Containment Integrity Program of OP-923. There is NO intent to dispatch an operator to locally check these valves. If they are open, an operator will be pre-designated for closure in the event of an accident. Directions have been provided in the RNO for isolating instrument air to a failed open S/G PORV. A note has been included reminding the operator that at least one PORV is needed for decay heat removal if the condenser is not available. Thus for a ruptured S/G one would not wish to use the common isolation valve which isolates motive air to all three S/Gs if the condenser is unavailable since there would then be no way to cool down the RCS. The final section of the procedure deals with minimization of the spread of secondary contamination. These actions are consistent with those performed in Path-2 and AOP-035. The actions are credited in the response to INPO SOER 93-01 GS-36 is closed to assure that the Auxiliary Boilers will be capable of supplying Gland Seal Steam. (OPEX AR 18118, Indian Point 2 SGTR Event) Supplement H Supplement H is intended to be used during a Loss of DC Bus A or B to verify Containment Isolation. Loss of DC will remove power to numerous lights for the Status Panels (Pink & Blue Lights). The Supplements are constructed to check only those valves that do not have redundant indications on the panels or the RTGB. Thus, the Operator would check all indications that are available via normal mechanisms, and in addition, those valves listed in the appropriate section of Supplement H. On a loss of DC Bus A (step 1), only one valve does not have redundant indication (SI-855). On a loss of DC Bus B there are multiple valves without redundant indication. It is expected that checking these valves locally will take some time. This is considered acceptable for several reasons. The accident in question is beyond the design basis for the plant, and is therefore not considered to be a credible accident. The valves all fail on a loss of power to the close position. Finally, the procedure assumes NO other accident in progress for a loss of DC. I EPP-SUPP-BD Rev. 35 Page 7 of 10 I HLC-08 NRC Written Exam 9. Given the following: -The plant was operating at 100% RTP when S/G "A" Steam line severs inside Containment. Which ONE (1) of the following describes the FW valves that should have CLOSED? A. FRVs for ALL THREE S/Gs (FCV-478, 488, 498) FRV BYPs for ALL THREE S/Gs (FCV-479, 489, 499) FW HDR SECTION Valves for ALL THREE S/Gs (V2-6A1B/C) B. FRVs for ALL THREE S/Gs (FCV-478, 488, 498) FRV BYPs for ALL THREE S/Gs (FCV-479, 489, 499) FW HDR SECTION Valve for S/G "A" ONLY (V2-6A) C. FRV for S/G "A" ONLY (FCV-478) FRV BYP for S/G "A" ONLY (FCV-479) FW HDR SECTION Valve for S/G "A" ONLY (V2-6A) D. FRV for S/G "A" ONLY (FCV-478) FRV BYP for S/G "A" ONLY (FCV-479) FW HDR SECTION Valves for ALL THREE S/Gs (V2-6A1B/C) 9 \11) I s kl-y?' Hew wocJc\t+ S.\&v1se b PR-y d: r-{<v 1S/p+o dose-OV] 1y fe/l-iJvce:. "¥i" ':1 e-+ r=\iYf'o iSo\c<,.:\-.-J t<J 0\-\ \ "3 <;/6r5? HLC-08 NRC Written Exam 9. 000040 AAl.02 OOllSTM LINE RUPTURE/1I1I4.5/4.5IROIHIGHININNEW -2008/ESF-005 Given the following: -The plant was operating at 100% RTP when S/G "A" Steam line severs inside Containment. Which ONE (1) of the following describes the FW valves that should have CLOSED? A:I FRVs for ALL THREE S/Gs (FCV-478, 488, 498) FRV BYPs for ALL THREE S/Gs (FCV-479, 489, 499) FW HDR SECTION Valves for ALL THREE S/Gs (V2-6A/B/C) B .. FRVs for ALL THREE S/Gs (FCV-478, 488, 498) FRV BYPs for ALL THREE S/Gs (FCV-479, 489, 499) FW HDR SECTION Valve for S/G "A" ONLY (V2-6A) C. FRV for S/G "A" ONLY (FCV-478) FRV BYP for S/G "A" ONLY (FCV-479) FW HDR SECTION Valve for S/G "A" ONLY (V2-6A) D. FRV for S/G "A" ONLY (FCV-478) FRV BYP for S/G "A" ONLY (FCV-479) FW HDR SECTION Valves for ALL THREE S/Gs (V2-6A/B/C) The correct answer is A. A: Correct -A S/G steam line severing inside Containment would cause CV pressure to increase to above the CV SI setpoint. When Safety Injection actuates, feedwater will be isolated to ALL S/Gs. B: Incorrect -Safety Injection isolates ALL FW HDR Section valves. C: Incorrect -Correct answer for FRV and FRV Bypass valves if there were a high S/G level. FW HDR Section valves do NOT go CLOSED on a S/G high level. D: Incorrect -Correct answer for FRV and FRV Bypass valves if there were a high S/G level. FW HDR Section valves do NOT go CLOSED on a S/G high level. Exam Question Number: 9

Reference:

SD-006, ESF, Figure 12.SD-027, FW, Page 20. KA Statement: Ability to operate and / or monitor the following as they apply to the Steam Line Rupture: Feedwater isolation. History: New -Written for HLC-08 NRC Exam. Tuesday, June 17,20081:21:02 PM 11 I Not Redundant --.j I I L-CLOSE , (5 SEC) FEEDWATER ISOLATION LOGIC ESF-FIGURE-12 STMGEN.A HIGH LEVEL n m CLOSE CLOSE .... R (5 SEC) BYPASS FEEDWATER _____ --,.1 CONTROL VAlVE MAIN FEEDWATER FCv-479 STMGEN. B HIGH LEVEL CLOSE CLOSE CLOSE (5 SEq (5 SEC) .I BYPASS FEEDWATER MAIN FEEDWATER CONTROL VALVE I STMGEN.C HIGH LEVEL n m Nat Redundan'r- __ I I CLOSE CLOSE CLOSE , (5 SEC) * (5 SEC)/ MAIN FEEDWATER FCV-499 _____ , CONTROL VALVE FCV-488 FCV-489 v CONTROL VAlVE FCV-498 vr-------------------J/ STM. GEN. A STMGEN B MAIN FEEDWATER PUMP TRIP INFORMATION USE ONLY TURBINE TRIP STMGEN.C SD-027 F eedwater System program level. 2. Depress the AUTO pushbutton on the FRV controller AND slowly close the respective bypass valve. 3. Verify each FRV in AUTO is maintaining programmed level. 4. If necessary a FRV can be placed in Manual to control level. The feedwater pump recirculating valve closes as flow is increased above its set point of 3100 gpm. The second condensate pump is started at 45-50 % power and the second main feedwater pump is started at 55-60% power. As the unit is shutdown, the above procedure is followed in reverse order. 6.2 Abnormal Operation AOP-010, Main Feedwater/Condensate Malfunction AOP-025, RTGB Instrument Failure 6.2.1 Feedwater flow will be affected by receipt of the following signals in the manner described: FW 1) Low TAVG (554 OF) plus a reactor trip will close the steam generator feedwater regulating valves, FCV-478, FCV-488, and FCV-498. 2) A high steam generator level (75% narrow range) in any steam generator will close the associated feedwater regulating valve, FCV -478, FCV -488, or FCV-498, the associated feedwater regUlating bypass valve, FCV-479, FCV-489, or FCV -499, trip the main feedwater pumps and trip the turbine. 3) A steam/feedwater flow mismatch (0.64 X 10 6 lbs/hr) in conjunction with a low steam generator level (30 % narrow range) will initiate a reactor trip. 4) A steam generator low-low water level (16% narrow range) will initiate a reactor trip and start the auxiliary feedwater pumps. :c:. ---------- .. -.--.------------

5) A Safety Injection signal will close the feedwater header section valves, V2-6A, FW-V2-6B, and FW-V2-6C, open the feedwater pump breakers, Close the steam generator regulating valvesj'CV-478JCV-488, the steam generator feedwater FCV-479, FCV-489, and FCV-A99.-_

6) An AMSAC signal, if not bypassed, will cause a turbine trip;the--atmliary feedwater pumps to start, and the steam generator blowdown valves to close. Page 20 of 35 Revision 8 INFORMATION USE ONLY HLC-08 NRC Written Exam 10. Given the following:

-The Reactor is at 4% RTP in preparation for Turbine startup. -Main Feedwater Pump "A" is under clearance for maintenance. -Main Feedwater Pump "8" is operating. -AFW Pumps are shutdown and aligned for automatic operation. -AMSAC is aligned in NORMAL. -Narrow range Steam Generator levels are at 44%. -Steam Dumps indicate throttled OPEN. Which ONE (1) of the following statements describes the AFW Pump status immediately after Main Feedwater Pump "8" trips? A. The MDAFW and SDAFW Pumps must be manually started. B. The MDAFW Pumps have auto started but the SDAFW Pump must be manually started. C. The SDAFW Pump has auto started but the MDAFW Pumps must be manually started. D. The MDAFW and SDAFW Pumps have auto started. \1l v ' L. '--\.> \(\tr\-iP 0 . .1..6; \)\'.:: . O .. 'rio 0*\-\,,<1,1-CeMd.'B tM-.S w& u>.d Q....\r-Q..\ S\Jf-t-PW VV"t\A.p Sl-o.,....\-(N--r) c\ '?(}Vv<.f.5 10 A-s s-k\ .. d '<\ ,\--\l <;;,+<.M.\: ,. t1-\-=l<.,.) Y' CtAA:\n o CLtl tNt * 'I ]: -\-\ 5 obv;"v.s c",'r kct :,'\-. G-f- \..'/1 \\ OV\ 0" 105.5 ok'- Nk Pb,M.G,h!-. HLC-OB NRC Written Exam 10.000054 AAl.02 OOlfLOSS OF MFW/1I1I4.4/4.4IROIHIGH/N/A/COOK -2002/AFW-008 Given the following: -The Reactor is at 4% RTP in preparation for Turbine startup. -Main Feedwater Pump "A" is under clearance for maintenance. -Main Feedwater Pump "B" is operating. -AFW Pumps are shutdown and aligned for automatic operation. -AMSAC is aligned in NORMAL. -Narrow range Steam Generator levels are at 44%. -Steam Dumps indicate throttled OPEN. Which ONE (1) of the following statements describes the AFW Pump status immediately after Main Feedwater Pump "B" trips? A. The MDAFW and SDAFW Pumps must be manually started. B:t The MDAFW Pumps have auto started but the SDAFW Pump must be manually started. C. The SDAFW Pump has auto started but the MDAFW Pumps must be manually started. D. The MDAFW and SDAFW Pumps have auto started. The correct answer is B. A: Incorrect -Both MDAFW Pumps start on loss of Main FW, both MFP breakers being OPEN is a Loss of Feedwater. B: Correct -The MDAFW Pumps will auto start on the loss of both Main FW Pumps but the SDAFW Pump will NOT. C: Incorrect -The SDAFW will only start on Lo-Lo S/G level, AMSAC, and MFP Bus UV. MDAFW Pumps will start on Lo-Lo Level, AMSAC, loss of Main FW, Blackout, & SI. D: Incorrect -SDAFW will NOT auto start. AMSAC will NOT ARM until power has been raised to> 35%. Exam Question Number: 10

Reference:

SD-042; AFW; Pump Start Logics, Figure 11. KA Statement: Ability to operate and / or monitor the following as they apply to the Loss of Main Feedwater (MFW): Manual startup of electric and steam-driven AFW pumps. History: Changed correct answer to B from D. Tuesday, June 17, 2008 1 :21 :02 PM 12 AFW PUMP AUTO-START LOGIC AFW-FIGURE-ll SG "A" 2/3 SG "B" 2/3 SG "e" 2/3 LOW lOW lOW LOW LOW lOW lEVEL lEVEl lEVEL 1/3 SG lO-lO lEVEL AUTO-START BLOCK IN MAIN FEED PUMP BRKRS NORMAL OPEN ....... A B MFW BKRS OPEN AUTO-START BLOCK IN NORMAL LOSS OF POWER El/E2 OR EDG BRKRS CLOSED SAFEGUARDS AND BLACKOUT SEQUENCES MANUAL I I 2/3 SG lO-lO lEVEL AUTO-START BLOCK IN NORMAL lOSS OF POWER 4 kV BUS BUS 1 BUS 4 UV UV UV UV _15SEC,_ T/D MANUAL START MDAFW A&B OPEN MDAFW PUMP CLOSE SGBD OPEN TURBINE START TURBINE DRIVEN PUMP (OPENS Vl-8A, B, C) DISCHARGE VALVES ISOLATION VALVES PUMP DISCHARGE (V2-16A, B, C) VALVES (V2-14A, B, C) INFORMATION USE ONLY Cook Nuclear Plant: Reactor Operator License Exam December 2002 28. 0361 Unit 1 Reactor is at 4% power in preparation for Turbine startup with the following conditions: -West Main Feedwater pump is tripped. -East Main Feedwater pump is operating. -AFW pumps are shutdown and aligned for automatic operation. -AMSAC is aligned in NORMAL. -Narrow range steam generator levels are now 44%. -Steam Dumps indicate 8% open. Which ONE of the following statements correctly describes the AFW pump status immediately after the East Main Feedwater Pump trips? a. The Motor Driven and Turbine Driven AFW pumps must be manually started. b. The Motor Driven AFW Pumps have auto started but the Turbine Driven AFW pump must be manually started. c. The Turbine Driven AFW Pump has auto started but the Motor Driven AFW pumps must be manually started. d. The Motor Driven and Turbine Driven AFW pumps have auto started. Lesson Plan/Obj: RO-C-05600 / #15

Reference:

01-0HP-4022-055-001, Loss of Main FW Pump; 01-0HP-4021-001-006, Power Escalation; SOD -05600-001, Auxiliary Feed System Loss of Main Feedwater (MFW) -Ability to operate and/or monitor the following as they apply to the Loss of Main Feedwater (MFW): Manual startup of electric and steam-driven AFW pumps Answer: 8 HLC-08 NRC Written Exam 11. During a Loss of All AC Power event, EPP-1, LOSS OF ALL AC POWER directs the isolation of seal cooling for events of 15 minutes or more without RCP seal cooling. Which ONE (1) of the following is the reason for isolating seal cooling? A. RCP seals and shafts may be damaged by thermal shock when power is restored to the OS Bus. B. RCP thermal barriers are susceptible to rupture from thermal shock when power is restored to the OS Bus. C. Steam binding of the Charging Pumps from RCP seal leakoff flashing when a Charging Pump is restarted. D. RCP Thermal Barriers are susceptible to rupture from water hammer when power is restored to the OS Bus. 11 HLC-08 NRC Written Exam 11. 000056 AA2.67 OOl/LOSS OF OFFSITE POWRIll1l2.9/3.1fRO/LOWfN/NNEW -200S/EPP-I-004 During a Loss of All AC Power event, EPP-1, LOSS OF ALL AC POWER directs the isolation of seal cooling for events of 15 minutes or more without RCP seal cooling. Which ONE (1) of the following is the reason for isolating seal cooling? A'I RCP seals and shafts may be damaged by thermal shock when power is restored to the OS Bus. B. RCP thermal barriers are susceptible to rupture from thermal shock when power is restored to the OS Bus. C. Steam binding of the Charging Pumps from RCP sealleakoff flashing when a Charging Pump is restarted. D. RCP Thermal Barriers are susceptible to rupture from water hammer when power is restored to the OS Bus. The correct answer is A. A: Correct -Injection loss lasting> 15 minutes may allow seals and shaft to heatup to RCS temperature. Engineering and manufacturer will be consulted prior to restoration. B: Incorrect -It is the seal themselves (NOT the Thermal Barrier) which are susceptible to failure from thermal shock. C: Incorrect -Steam binding is a possibility due to no CCW, but NOT a reason for isolation after a loss of > 15 minutes. 0: Incorrect -Water hammer may occur if voids or steam formation in seal injection lines occurs, but it is NOT a reason for isolation after a loss of > 15 minutes. Exam Question Number: 11

Reference:

EPP-1 Page 5; EPP-1 BO, Pages 4-6,64. KA Statement: Ability to determine and interpret the following as they apply to the Loss of Offsite Power: Seal injection flow (for the RCPs). History: New -Written for HLC-08 NRC Exam. Tuesday, June 17, 2008 1 :21 :03 PM 13 Rev. 38 EPP-l LOSS OF ALL AC POWER Page 5 of 61 INSTRUCTIONS RESPONSE NOT OBTAINED * * * * * * * * *

• * * * * * * ..
• "f'Y"k ."""':':-t.
• -;-;"* * * * * * * *
• * . CAUTION ------.-IF more than 15 minutes elapses without RCP Seal Cooling, THEN Seal Cooling (CCW AND Seal Injection)

MUST be isolated before starting CCW OR Charging or Seal Damage could occur. *************************************************************************

• 4. Check Seal Cooling From CCW As Follows: a. Check CCW Pump "A" RUNNING b. Go To Step 6 Perform The Following To Isolate RCP Seal Cooling: a. In Pipe Alley close FCV-626, THERM BARRIER OUTLET b. In the Charging Pump Room, close the RCP SEAL WATER FLOW CONTROL VALVEs
• CVC-297A
• CVC-297B
• CVC-297C a. IF RCP Seal Cooling is NOT OR can NOT be restored in less than 15 minutes, THEN perform Step 5. Go To Step 6. a. Close CC-736,CC FROM RCP "A", "B", "c" THERMAL BARRIER.

2. DESCRIPTION A total loss of ac power at a nuclear power station can result only through a coincident loss of grid power from the high voltage distribution lines serving the station and some combination of events preventing the station emergency diesel generators from energizing the emergency ac busses. The immediate consequences of the loss of ac power, if not accompanied by some other complicating event such as a loss of reactor coolant, loss of secondary coolant or steam generator tube rupture, are not severe. However, should ac power either from the grid or the emergency diesels not be restored quickly, the consequences to plant and public safety can potentially be extreme. The degree of severity of a loss of all ac power depends primarily on the duration of the ac power outage and the response of the reactor coolant pump (RCP) shaft seals to the loss of seal cooling, i.e., the simultaneous loss of high pressure injection flow to the RCP seals and of component cooling water (CCW) flow to the RCP thermal barrier. Loss of high pressure seal injection flow from the motor driven charging/SI pumps results in out leakage from the RCS along the RCP shafts. Without power this leakage cannot be replaced and a continuous loss of reactor coolant occurs with time. Loss of RCP seal cooling potentially can also cause degradation of the sealing capability of the RCP seals as a result of overheating.

Degradation of the sealing capability may result in an increase in leakage out of the RCS from several gpm per RCP to several hundred gpm per RCP. To mitigate the severity of a loss of all ac power, it is necessary to minimize RCS inventory loss with time and to restore ac power so that RCS inventory can be restored. Various aspects of the loss of all ac power event are discussed as appropriate in succeeding subsections. Due to the key role of the RCP following the loss of all ac power, a more detailed RCP description is given in the following subsection. Additional information is contained in WCAP-1 0541, Revision 2, Westinghouse Owners Group Report, "Reactor Coolant Pump Seal Performance Following a Loss of All AC Power" (Reference 4). 2.1 RCP Description RCP Seal System Leakage The RCP is designed such that leakage along the RCP shaft is controlled by three shaft seals arranged in series. Under normal operating conditions, the RCP shaft seal system functions to control and direct RCP seal leakage such that leakage to the containment is essentially zero. The following briefly describes the three shaft seals. o Number 1 Seal The number 1 seal is the main seal. It is a controlled leakage film riding seal primarily consisting of a runner which rotates with the shaft and a non-rotating seal ring attached to the seal housing. The flow path is between the faceplate interface of the seal ring and seal runner, with leakage rate depending on faceplate taper angle and separation. The number 1 seal is designed to accommodate full RCS pressure of 2235 psig and limit RCS nominal leakage to approximately 3 gpm, with a maximum acceptable leakage of approximately 4.5 gpm. The major portion of the number 1 seal leakage is directed to the number 1 sealleakoff line with the smaller portion directed through the number 2 seal. The number 1 sealleakoff line normally returns leakoff through containment to the charging/SI pump suction. This seal return line includes a relief valve inside containment which directs leakoff flow to the pressurizer relief tank when containment is isolated. The relief valve has a nominal 150 psig set pressure. I EPP-1-BD Rev. 38 Page 4 of 1061 o Number 2 Seal The number 2 seal is a rubbing face seal consisting of a seal ring with shrink fit insert which rubs on a runner rotating with the shaft. With the number 1 seal operative, the number 2 seal is designed to accommodate a nominal differential pressure of 30 psig and limit RCS leakage to approximately 3 gallons per hour. If the number 1 seal becomes inoperative, the number 2 seal is designed to accommodate full RCS pressure of 2235 psig and limit RCS leakage to approximately 8 gpm, with a maximum acceptable leakage of approximately 12 gpm. The number 2 seal functions to create a backpressure on the number 1 seal, thus directing the major portion of number 1 seal flow to the number 1 sealleakoff line. The major portion of leakage through the number 2 seal is directed to the number 2 sealleakoff line with a small portion directed through the number 3 seal, depending on number 3 seal design. o Number 3 Seal The number 3 seal is a rubbing-face seal that functions to limit leakage to the containment environment under normal pump operating conditions. It is designed to accommodate low differential pressures with number 3 seal leakoff being directed to the reactor coolant drain tank or containment sump, depending on number 3 seal design. The number 3 seal is not a pressure boundary seal and does not play an important role in limiting RCP seal leakage following the loss of all ac power. Under normal operating conditions, the controlled leakage shaft seal system is cooled by independent and redundant cooling systems. Adequate seal cooling for continued RCP operation can be provided by either high pressure seal injection flow from the charging/SI pumps or low pressure CCW flow circulated through the RCP thermal barrier. Seal injection flow acts as a buffer to prevent reactor coolant from entering the pump seal and bearing section of the pump. A portion of seal injection flows down the shaft and into the RCS while the remainder flows up through the seal system. If seal injection to the RCP is lost, the hot reactor coolant can flow up the pump shaft. Under this condition, the RCP thermal barrier functions as a heat exchanger to cool the hot reactor coolant before it enters the RCP bearing and seal area. The RCP is designed to accommodate the temporary loss of seal injection flow and CCW flow that accompanies a loss of offsite power, including the normal time delays associated with reestablishing these RCP support systems on emergency ac power. This is accomplished by the volume of cool water in the seal area and the time that it takes to leak this water through the RCP seals prior to hot water entering the seals. The RCP is designed to accommodate the loss of support systems for one minute following loss of offsite power. Under best estimate conditions, the RCP design should preclude hot water from entering the seal area for several minutes. Under the loss of all ac power event, RCP support systems may not be restored prior to the introduction of hot reactor coolant into the seal system. Under this condition, the RCP seal leak rate becomes dependent on RCS temperature as well as the RCS pressure. At temperatures in excess of 300°F, RCP seal system sealing capability and sealing life may start to degrade with consequential increase in seal leakage flow. The potential for degradation in sealing capability and sealing life increases with increasing temperature above 300°F. Seal performance under high temperature conditions is difficult to analyze due to several interacting considerations, including:

1) The seal a-ring material softens with increasing temperature, affecting the a-ring sealing ability and life. 2) The thermal gradients affect the faceplate tapers of the number 1 seal ring and runner and the shrink fit of the number 2 seal ring insert, affecting sealing surfaces.

Nonuniform thermal gradients and extrusion of a-rings may result in nonuniform sealing surfaces.

3) The leakage of reactor coolant through the seals could result in crud blockage of the seals. I EPP-1-BD Rev. 38 Page 5 of 1061 Due to the several interacting considerations, it is difficult to analytically or experimentally predict the RCP leakage flow following a loss of all ac power. Further, it is possible that different RCPs may be affected in different ways with resulting variations in leakage flow. Since it is difficult to accurately predict RCP seal leakage flow following a loss of all ac power, field experience has been reviewed for relevant information and an analysis has been performed to estimate a maximum leakage rate. Although some field experience exists wherein both seal injection flow and thermal barrier CCW flow have been lost, it is of such a limited nature that RCP seal behavior cannot be accurately predicted.

General conclusions drawn from this experience indicate that seal leakoff flow can be expected to increase above the normal range during the loss of all ac power event, likely going above the seal leakoff flow instrument typical upper range of 6 gpm. However, based on this limited experience, abnormally excessive leakage rates were not experienced prior to restoration of RCP support systems. To evaluate the most severe consequences of a loss of all ac power to the RCP seal system, a conservative maximum RCP leakage rate is estimated to be 300 gpm. This rate is estimated by assuming that total RCS pressure of 2235 psig exists across the Rep thermal barrier labyrinth seals with the controlled leakage seals totally ineffective in controlling leakage flow. This estimate does not include credit for floating ring seals which are in a limited number of RCPs. Maximum leakage rates for RCPs with floating ring seals will be less than 300 gpm. Based on the above discussion, RCP seal integrity is a major concern during a loss of all ac power event. The high RCS temperatures and pressures characteristic of a plant no-load condition can lead to eventual RCP seal degradation and increased RCS inventory loss. This seal degradation can be mitigated by reducing the RCS pressure and temperature consistent with other plant constraints. Reducing RCS pressure reduces leakage flow through the RCP seals, thereby reducing RCS inventory loss for a given seal condition. Reducing RCS temperature reduces the thermal degradation of materials and thermal expansion effects that tend to degrade the seal system sealing capability and sealing life. Consequently, any actions to reduce RCS pressure and temperature during a loss of all ac power event are consistent with minimizing RCS inventory loss and maximizing time to core uncovery. Benefits and Consequences of Restoring Seal Cooling Following the restoration of ac power, the operator will have the capability to restore seal cooling by reestablishing seal injection flow or reestablishing thermal barrier cooling using the component cooling water system. Restoring seal cooling may have several benefits such as reducing seal leakage and preventing further damage to the seal components. However, Westinghouse has not performed an analysis of how the RCP seal package will react as the seals cool, fits contract, the shaft moves, etc., possibly with partially extruded O-rings. There may be a potential to make seal leakage worse by restoring seal cooling, depending on how it is done. I The RCP Vendor Manual for...reestablishing seal cooling to a hot seal to prevent further damage due to thermal shock and to prevent warping of the RCP shaft due to uneven cooling. These limits are only intended for a loss of seal cooling of short enough duration that the seal package heatup is limited. Although the limits have been extrapolated for an extended loss of seal cooling event in the past, they have not been validated for such an event that is beyond the design basis of the RCP. Therefore, no specific conclusions may be taken from the RCP vendor manual guidance for reestablishing seal cooling following an extended loss of seal cooling event. The following provides a qualitative assessment that determines the most appropriate method of restoring seal cooling following an extended loss of all ac power event: I EPP-1-BD Rev. 38 Page 6 of 1061 I EPP-1-BD WOG BASIS/DIFFERENCES STEP 8 WOG BASIS PURPOSE: To isolate the RCP seals BASIS: (Updated for Rev 2 ERG) This step groups three actions, with different purposes, aimed at isolating the RCP seals. The actions are grouped since all require an auxiliary operator, dispatched from the control room, to locally close containment isolation valves (the reference plant utilizes motor operated valves for the RCP seal return, RCP thermal barrier CCW return lines and RCP seal injection lines). This grouping assumes that the subject valves are located in the same penetration room area and that they are accessible. Concurrent with dispatching the auxiliary operator, the control room operator should place the valve switches for the motor operated valves in the closed position 50 that the valves remain closed when ac power is restored. Isolating the seal return line prevents seal leakage from filling the volume control tank (VCT) (via seal return relief valve outside containment) and subsequent transfer to other auxiliary building holdup tanks (via VCT relief valve) with the potential for radioactive release within the auxiliary building. Such a release, without auxiliary building ventilation available, could limit personnel access for local operations .

• Isolating the RCP seal injection lines preeares the plant for recovery IIibila.protecting from seal and shaft damageJha1..ala.Y.QQQ!,Ir when a charging PUInR is started as part of the recovery.

With the RCP seal injection lines isolated, a charging pump can be started in the normal charging mode without concern for cold seal injection flow thermally shocking the RCPs. Seal injection can subsequently by established to the RCP consistent with appropriate plant specific procedures. Isolating the RCP thermal barrier CCW return outside containment isolation valve prepares the plant for recovery while protecting the CCW system from steam formation due to RCP thermal barrier heating. Following the loss of all ac power, hot reactor coolant will gradually replace the normally cool seal injection water in the RCP seal area. As the hot reactor coolant leaks up the shaft, the water in the thermal barrier will heat up and potentially form steam in the thermal barrier and in the CCW lines adjacent to the thermal barrier. Subsequent automatic start of the CCW pump would deliver CCW flow to the thermal barrier, flushing the steam into the CCW system. If abnormal RCP seal leakage had developed in a pump, the abnormally high leakage rate could exceed the cooling capacity of the CCW flow to that pump thermal barrier and tend to generate more steam in the RCP thermal barrier CCW return lines. Isolating these lines prevents the potential introduction of this steam into the main portion of the CCW system upon CCW pump start. This keeps the main portion of the CCW system available for cooling equipment necessary for recovering the plant when ac power is restored. KNOWLEDGE: RCP seal integrity concerns following loss of ac power (See Subsection 2.1 ) RNP DIFFERENCES/REASONS There are essentially no differences other than it is the expectation of the ERG that step will always be implemented while the RNP step is only expected to be needed for beyond license basis events. SSD DETERMINATION This is not an SSD. Rev. 38 Page 64 of 106/ HLC-08 NRC Written Exam 12. Given the following: -The plant is in MODE 6 with refueling in progress. -Source Range Channel N-31 is INOPERABLE. -BOTH PAM Source Range monitors (N-51A and N-52A) are OPERABLE. -Instrument Bus 2 trips due to a fault on the bus. . -\ ( .. '1 1: ,,' L.-,. . \-.,. /l; '1 If ,/, p.,..J Y (}I' II .. :*\:' 5"'\ ..... ) '1<t \......:.,tl. .I\\I\lr) Which ONE (1) of the following LCO REQUIRED ACTIONS is applicable? A. Verify one PAM Source Range monitor provides indication in the CR within 15 minutes AND Log indicated SR count rate every 30 minutes. B. Verify one PAM Source Range monitor provides indication in the CR within 15 minutes, suspend core alterations AND suspend positive reactivity additions. C. IMMEDIATELY initiate action to restore one SR monitor to OPERABLE AND suspend core alterations AND suspend positive reactivity additions. D. IMMEDIATELY initiate action to restore one SR monitor to OPERABLE AND log indicated SR count rate every 30 minutes. 12 HLC-08 NRC Written Exam 12.000057 G2.2.22 OOllLOSS OF VITAL AC INSI1I1I4.0/4.7IROIHIGH/N/NNEW -200S/GP-01O-002 Given the following: -The plant is in MODE 6 with refueling in progress. -Source Range Channel N-31 is INOPERABLE. -BOTH PAM Source Range monitors (N-51A and N-52A) are OPERABLE. -Instrument Bus 2 trips due to a fault on the bus. Which ONE (1) of the following LCO REQUIRED ACTIONS is applicable? A. Verify one PAM Source Range monitor provides indication in the CR within 15 minutes AND Log indicated SR count rate every 30 minutes. B. Verify one PAM Source Range monitor provides indication in the CR within 15 minutes, suspend core alterations AND suspend positive reactivity additions. IMMEDIATELY initiate action to restore one SR monitor to OPERABLE AND suspend core alterations AND suspend positive reactivity additions. D. IMMEDIATELY initiate action to restore one SR monitor to OPERABLE AND log indicated SR count rate every 30 minutes. The correct answer is C. A: Incorrect -LCO for ONE SR inoperable (Condition A). B: Incorrect -LCO for ONE SR inoperable AND completion times of Condition A not met. C: Correct -Loss of the instrument bus results in BOTH SR monitors inoperable. This requires LCO 3.9.2, Condition C to be implemented. 0: Incorrect -Correct statement for 2 SR INOPERABLE, but logging SR count rate assumes that NI-51 AlNI-52A Gammametrics can be substituted for SR monitors. Exam Question Number: 12

Reference:

ITS 3.9.2; GP-01 0, Pages 11 and 56. KA Statement: Knowledge of limiting conditions for operations and safety limits. History: New -Written for HLC-08 NRC Exam. Tuesday, June 17, 2008 1 :21 :03 PM 14 --,--;.,----- 3.9 REFUELING OPERATIONS Nuclear Instrumentation 3.9.2 3.9.2 Nuclear Instrumentation LCO 3.9.2 Two source range neutron flux monitors shall be OPERABLE. APPLICABILITY: MODE 6. ACTIONS CONDITION A. One required source range neutron flux monitor inoperable HBRSEP Unit No. 2 REQUIRED ACTION COMPLETION TIME A.I Verify one Post 15 minutes Accident Monitor (PAM) source range neutron flux monitor provides indication in the Control Room. AND. A.2 Log indicated PAM 30 minutes source range neutron monitor count rate. AMD. 3.9*2 Once per 30 minutes thereafter (continued) Amendment No. tT6,tBfr,190 .J b ACTIONS (continued) CONDITION B. Required Actions and B.1 Completion Times of Condition A not met. 00 B.2 C. Two required source C.1 range neutron fl ux monitors inoperable. 00 C.2 00 C.3 00 C.4 HBRSEP Unit No. 2 Nuclear Instrumentation 3.9.2 REQUIRED ACTION COMPLETION TIME Suspend CORE ALTERATIONS. Immediately Suspend operations that would cause Immediately introduction into the RCS. coolant with boron concentration less than required to meet boron concentration of LCO 3.9.1. Initiate action to Immediately restore* one source range neutron flux monitor to OPERABLE status. Suspend CORE Immediately ALTERATIONS. Suspend positive reactivity additions. Immediately Perform SR 3.9.1.1. 4 hours AND. Once per 12 hours thereafter 3.9-3 Amendment No. tT6.t86.190 5.18 The principles of ALARA shall be used in planning and performing work and operations in the Radiation Control Area. 5.19 Materials used to construct the RC-FMEA and Transfer Canal Cover may be considered transient combustibles and shall be handled lAW FP-003. 5.20 A case evaluation has been performed for each section of this procedure lAW PLP-037. The case determination is defined at the beginning of each appropriate section or evolution. 5.21 Proper lighting shall be provided in the area of fuel handling operation. If anyone involved in the fuel handling operation is not satisfied with the lighting, then additional temporary lights should be installed. 5.22 ITS LCO 3.9.6 requires that Refueling Cavity water level be greater than or equal to 23 feet above the top of the Reactor Vessel flange. Refueling Cavity Water level is measured from the operating deck down. 8ased on UFSAR Figure 5.3.0-1 and drawing G-190188, 23 feet above the Reactor Vessel flange corresponds to 35 inches below the operating deck. Maintaining level higher . than 29 inches below the operating deck ensures water level is at least 6 inches above the required ITS level. 5.23 ITS LCO 3.7.12 requires that SFP level be verified greater than or equal to 21 feet above the top of irradiated fuel assemblies seated in the storage racks. SFP level is measured from the SFP floor up. According to the UFSAR the top of the fuel is located 14 feet, 1/4 inch above the SFP floor. Maintaining level above 35 feet, 1/4 inch satisfies the required ITS level. 5.24 When the SFP GATE VALVE is open, it takes 1500 gallons of makeup to raise level 1 inch in the Spent Fuel Pool and Refueling Cavity. 5.25 The Fuel Transfer Cart shall be on the CV side at the end stop position before the SFP GATE VALVE is closed. [CAPR NCR 00239329/NCR 00231270] 5.27 ITS 3.9.2 requires two Source Range Neutron Flux Monitors to be operable in MODE 6. If one of the 8F3 Source Range Detectors becomes inoperable, a PAM Source Range Detector may be used if the requirements of ITS LCO 3.9.2A are met. RCP "8" and RCP "c" should NOT be uncoupled at the same time if the RHR System is required to be operable AND RCS level is at or below -50 inches. Uncoupled RCPs result in inaccurate standpipe level indication when RCS level is at or below -50 inches. (ESR 95-00649) (NCR 30599) I GP-010 Rev. 66 Page 11 of 791 ATTACHMENT 10.2 Page 3 of 4 MODE 6 CHECKLIST

5. AC Instrument Bus source is operable to support the onsite AC Instrument Bus electrical power distribution subsystem(s) required by ITS LCO 3.8.10. 6. The necessary portion of AC, DC, and AC Instrument Bus electrical power distribution subsystems are operable to support equipment required to be operable. (ITS 3.8.10) 7. Boron concentration of the RCS, Refueling Canal and the Refueling Cavity is greater than or equal to the limits identified in COLR (FMP-001) for MODE 6 (ITS LCO 3.9.1) AND SFP boron greater than the limit specified for the RCS. RCS Boron ppm Refueling Canal ppm Refueling Cavity ppm FMP-001 ppm SFP ppm NOTE: ITS LCO 3.0.4 allows entry into MODE 6 with only one Source Range Neutron Flux Monitor operable provided the actions of ITS LCO 3.9.2 are met. A PAM Source Range Detector may be used in place of a BF3 detector if the requirements of ITS LCO 3.9.2 are met. V Two Source Range Neutron Flux Monitors are operable. (ITS LeO 3.9.2) __ I GP-010 Rev. 66 Page 56 of 791 HLC-08 NRC Written Exam 13. Given the following:

-The plant is operating at 100% RTP. -Two electricians are performing MST-902, "A" and "B" STATION BATTERY TEST, obtaining battery pilot cell temperatures, electrolyte levels and battery ground checks. -During the battery checks, Battery Charger "B" is tripped. Which ONE (1) of the following describes the effect on Battery Bus "B" and any ITS LCOs that are applicable? Battery Bus "B" is ... A. de-energized. An LCO is in effect to restore power to the bus. B. de-energized. LCO 3.0.3 is in effect. C. at -120VDC. An LCO is in effect to restore a battery charger to service. D. at -120VDC. NO LCO is in effect due to Battery Bus "B" energized within normal parameters. 13 HLC-08 NRC Written Exam 13. 000058 G2.2.36 OOllLOSS OF DC POWERlll1l3.1I4.2IRO/HIGH/NINNEW -2008/DC-01O Given the following: -The plant is operating at 100% RTP. -Two electricians are performing MST-902, "A" and "B" STATION BATTERY TEST, obtaining battery pilot cell temperatures, electrolyte levels and battery ground checks. -During the battery checks, Battery Charger "B" is tripped. Which ONE (1) of the following describes the effect on Battery Bus "B" and any ITS LCOs that are applicable? Battery Bus "B" is ... A. de-energized. An LCO is in effect to restore power to the bus. B. de-energized. LCO 3.0.3 is in effect. at -120VDC. An LCO is in effect to restore a battery charger to service. O. at -120VDC. NO LCO is in effect due to Battery Bus "B" energized within normal parameters. The correct answer is C. A: Incorrect -Bus B is still OPERABLE at 120 VDC, only the charger is tripped. Battery "B" is supplying the bus. B: Incorrect -Bus B is still energized from the battery. LCO 3.8.4 would be applicable, NOT 3.0.3 C: Correct -Voltage is still at 120VDC, only the charger has tripped. 2 hour LCO in effect. (3.8.4, Condition A) D: Incorrect -Voltage is still at 120VDC, only the charger has tripped. LCO 3.8.4 should be entered. This is the correct answer for MODE 5. MODE 5 does NOT require an LCO entry for loss of a battery charger if the Bus is energized. Exam Question Number: 13

Reference:

SO-038, DC Distribution; ITS 3.8.4; ITS 3.8.4 BD, Page B3.8-41. KA Statement: Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations. History: New -Written for HLC-08 NRC Exam. Tuesday, June 17, 20081:21:03 PM 15 SD-038 DC ELECTRICAL SYSTEM grounds; if the reading exceeds 10 volts either positive or negative the ground should be located and fixed as soon as possible. On Battery Chargers "A-I" and "B-1", the charger output voltmeter also serves as the DC ground meter. Ground indication is read by turning the selector switch to POS GND and then to NEG GND. Battery Charger "C" has a separate ground voltmeter to indicate DC ground in conjuction with the Ground VM switch. When not checking for ground, this voltmeter indicates the charger output. Ground indication is read by turning the selector switch to POS GROUND and then to NEG GROUND. During testing for grounds on Battery Chargers "A-1 ", "B-1 ", or "C", the voltmeter wilL

• dro to about 112 of buss voltage (approximatel 65 volts with no ound resent. A significantly lower rea mg on either direction would indicate a ground. If the reading drops to less than 50 volts, the ground should be located and fixed as soon as possible.

If a trouble alarm is being received on Battery Chargers "A-1 ", "B-1 ", or "C", I&C personnel can verify the ground is on the positive or negative buss by checking an LED indicator located inside the charger cabinet. Battery chargers "A -1" and "B-1" do not have a remote ground alarm test relay. Battery charger "C" has a test selector switch mounted below the ground indicator selector switch with POS GROUND and NEG GROUND positions to test the ground alarm. 4.1.1 DC Busses Each battery bus has indications for the bus voltage and also the amperes on the bus. The ammeters monitor current flow into or out of the battery attached to that MCC. 4.2 Alarms DC APP-036-Dl BATT CHARGER AlA-l TROUBLE, For battery Charger A this alarm is caused by, DC output fuse blown; DC overvoltage; AC input Failure; DC Ground; Ground test. For Charger A-I the alarm is caused by, DC overvoltage; AC input failure; DC output breaker open; charger failure; DC under voltage; DC ground. APP-036-D2 BATT CHARGER B/B-1 TROUBLE, For battery Charger B this alarm is caused by, DC output fuse blown; DC overvoltage; AC input Failure; DC Ground; Ground test. Page 12 of 18 Revision 6 INFORMATION USE ONLY ONE-LINE DIAGRAM OF "B" 125 VDC SYSTEM DC-FIGURE-2 (Rev. 0) BATTERY B M C C " BATTERY --CHARGER _,,_ MCC-6 "B" B INVERTER _,,_ B "" BATTERY --CHARGER B*1 DISTRIBUTION PANEL B PANEL B*A PANEL B*1 INFORMATION USE ONLY ". MCC-6 -3.8 ELECTRICAL POWER SYSTEMS 3.8.4 DC Sources -Operat i ng DC Sources -Operati ng 3.8.4 LCO 3.8.4 The Train A and Train B DC electrical power subsystems shall be OPERABLE. APPLICABILITY: MODES 1. 2. 3. and 4. ACTIONS CONDITION REQUIRED ACTION A. One DC electrical A.1 Restore DC electrical power subsystem power subsystem to inoperable. OPERABLE status. B. Required Action and B.1 Be in MODE 3. Associated Completion Time not met. AND B.2 Be in MODE 5. SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.8.4.1 Verify battery terminal voltage is 125.7 V on float charge. HBRSEP Unit No. 2 3.8*19 COMPLETION TIME 2 hours 6 hours 36 hours FREQUENCY 7 days (continued) Amendment No. 176 BASES APPLI CAB I LITY (continued) ACTIONS HBRSEP Unit No. 2 A.I DC Sources -Operating B 3.8.4 maintained in the event of a postulated DBA. The DC electrical pOWE!r requi rements for MODES 5 and 6 are addressed in the Bases for LCO 3.8.5. "DC Shutdown and During Movement of Irradiated Fuel Assemblies." Condition A represents one train with a loss of ability to completely respond to an event. and a potential loss of ability to remain energized during normal operation. It is. therefore, imperative that the operator's attention focus on stabilizing the unit. minimizing the potential for complete loss of DC power to the affected train. The 2 hour limit is consistent with the allowed time for an inoperable DC distribution system train. If one of the required DC electrical power subsystems is inoperable (e.g., inoperable battery, inoperable battery or battery charger and associated-- 1noperable battery). the remaining DC electrical power subsystem has the cilpacity to support a safe shutdown and to mitigate an accident condition. Since a subsequent worst case single active failure would. however, result in the complete loss of thE! remaining 125 VDC electrical power subsystems with attendant loss of ESF functions. continued power operation should not exceed 2 hours. The 2 hour Completion Time reflects a reasonable time to assess unit status as a function of the inoperable DC electrical power subsystem and, if the DC electrical power subsystem is not restored to OPERABLE status, to prepare to effect an orderly and safe unit shutdown. B.1 and 8.2 If the inoperable DC electrical power subsystem cannot be restored to OPERABLE status within the required Completion Time. the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the (continued) B 3.8*41 Revision No. 0 HLC-08 NRC Written Exam 14. Given the following: -The Unit was operating at 100% RTP, when a Pressurizer PORV failure caused a Reactor Trip. -APP-008-F7, SOUTH SW HDR LO PRESS, is received. -South SW header pressure is 29 PSIG and slowly decreasing. -North SW header pressure is 42 PSIG and stable. Which ONE (1) of the following contains the correct response of the Service Water system valves based on the above plant conditions? V6-16A, SW NORTH HEADER SUPPLY TO TURBINE BUILDING. V6-16B, SW SOUTH HEADER SUPPLY TO TURBINE BUILDING. V6-16C, SW ISOLATION TO TURBINE BUILDING. After ONE (1) minute with South SW header pressure less than ... A. 40 PSIG, ONLY valves V6-16B, and V6-16C will close. B. 40 PSIG, valves V6-16A, V6-16B and V6-16C will close. C. 31 PSIG, ONLY valves V6-16B and V6-16C will close. D. 31 PSIG, valves V6-16A, V6-16B and V6-16C will close. 14 HLC-08 NRC Written Exam 14. 000062 AK3.02 OOl/LOSS OF SERV WATER/l/l/3.6/3.9/ROIHIGH/N/AIRNP 2007/SW-008 Given the following: -The Unit was operating at 100% RTP, when a Pressurizer PORV failure caused a Reactor Trip. -APP-008-F7, SOUTH SW HDR LO PRESS, is received. -South SW header pressure is 29 PSIG and slowly decreasing. -North SW header pressure is 42 PSIG and stable. Which ONE (1) of the following contains the correct response of the Service Water system valves based on the above plant conditions? V6-16A, SW NORTH HEADER SUPPLY TO TURBINE BUILDING. V6-16B, SW SOUTH HEADER SUPPLY TO TURBINE BUILDING. V6-16C, SW ISOLATION TO TURBINE BUILDING. After ONE (1) minute with South SW header pressure less than ... A. 40 PSIG, ONLY valves V6-16B, and V6-16C will close. B. 40 PSIG, valves V6-16A, V6-16B and V6-16C will close. 31 PSIG, ONLY valves V6-16B and V6-16C will close. D. 31 PSIG, valves V6-16A, V6-16B and V6-16C will close. The correct answer is C. A: Incorrect -V6-16B and C will close. 40 PSIG setpoint is alarm setpoint for APP-008-F7 and incorrect for SW isolation. B: Incorrect -V6-16B and C will close, V6-16A will NOT because there is NO isolation signal for the North header. 40 PSIG setpoint is alarm setpoint for APP-008-F7 and incorrect for SW isolation. C: Correct -V6-16B and C will close at < 31 PSIG SW Header pressure coincident with a Turbine trip. D: Incorrect -V6-16B and C will close at < 31 PSIG SW Header pressure coincident with a Turbine trip. V6-16A will NOT close. 16 HLC-08 NRC Written Exam Exam Question Number: 14

Reference:

SO-004, SWS, Page 18, Figures 3,4; APP-008-F7. KA Statement: Knowledge of the reasons for the following responses as they apply to the Loss of Nuclear Service Water: The automatic actions (alignments) within the nuclear service water resulting from the actuation of the ESFAS. History: Changed Stem/answer to question valve realignment. (vs. SWB Pump status) Note: Repeat question from last RNP exam, Modified as listed in History. 17 SD-004 SERVICE WATER SYSTEM differential pressure is sensed across the screens. 4.5.4 PSL-1693A, B & C and FSL-1695A, B & C These switches are interlocked with the circulating water pump's starting circuits. These switches will prevent their respective circulating water pump from starting upon sensing either a low pressure or low flow in the SW lines supplying gland seal and bearing cooling water to the pumps. 4.5.5 TCV-1902A This valve is interlocked with the limit switches on the SDAFW pump's admission valves. The valve opens with the steam admission valves to provide cooling to the lube oil cooler. This valve is normally isolated (SW-246 locked closed). (The SDAFW self-cooling mode of operation is required due to the initial loss of SW during a station blackout event.) 4.5.6 TCV-1903A & B These valves are interlocked with the control circuits of MDAFW Pumps A and B to open on pump start and supply cooling water to the lube oil cooler. 4.5.7 V6-16A, B & C Auto closure feature to isolate Turbine Building:

• Valve V 6-16A will close if PSL-1616A reaches 31 psig decreasing for 60 seconds with a 20ET Turbine Trip signal present G Valve V6-16B will close if PSL-1684A reaches 31 psig decreasing for 60 seconds with a 20AST Turbine Trip signal present t;\ Valve V6-16C will close if PSL-1616B OR PSL-1684B reaches 31 psig l./ decreasing for 60 seconds with a 20ET OR 20AST Turbine Trip signal present Key lock switches (located in the Cable Spread Room) are used during maintenance, testing, or when the unit is in Cold Shutdown to inhibit the auto-close interlock.

4.5.8 SW Pump D sw Manual circuit breakers, with Kirk key interlock, are provided to power the pump from the 480v DS Bus in case of a fire. Page 18 of 35 Revision 11 INFORMATION USE ONLY SERVICE WATER PUMPS SW-FIGURE-3 HYPOCtLORITE --..... ... ux. BUILDING NORTH HEADER TOAux. BUILDING COMPONENTS V6*16A y, .168 AUX. BUILDING SOUTH HEADER AUX, BLDG * .-...".,., TURII. BlDG. SERVICE WATER TO TURBINE BUIWING COMPONENTS CIRCULATING WATER PUMPS V6*12A 1 ... l--...... STRAINER HYPOCHLORITE INFORMATION USE ONLY SERVICE WATER PUMPS 60 SEC m TURBINE BUILDING SW ISOLATION LOGIC (SIMPLIFIED) SW-FIGURE-4 TO TURBINE TRIP TO 20ET --I 1---20AST 31 PSIG DECREASING 60 SEC 1D .---VJJ..VE OPeN .--ViLVE OPEN INHIBIT ...--VAlVE OPEN ,--txt-INHIBIT CLOSE V6* 1 6A _-J---:::.:----==---CLOSE V6-16B . CLOSE V6-16 INFORMATION USE ONLY APP-008-F7 NOTE: Alarm may be received temporarily when shifting Service Water Pump due to system pressure surges, but should clear when Service Water System pressure stabilizes. No action is necessary in these instances. ALARM SOUTH SW HDR LO PRESS AUTOMATIC ACTIONS 1. None Applicable CAUSE 1. Loss of SW Pump(s) 2. CCW Heat exchanger Outlet Valves open too far 3. Rupture of Service Water Piping 4. Season increase in SW temperature (slow transient) OBSERVATIONS

1. Service Water Pressure (PI-1684, PI-1616) 2. Service Water Pump Breaker(s)

Indicating Lights ACTIONS 1. IF an operating SW Pump has tripped, THEN perform the following:

1) START a Standby Pump. 2) Dispatch operator to check breaker(s)

SW Pump A -480V Bus E1 (CMP 20B) SW Pump C -480V Bus E2 (CMP 24A) SW Pump B -480V Bus E1 (CMP 19C) SW Pump D -480V Bus E2 (CMP 25B) 3) Throttle CCW Heat Exchanger Return Valves, as necessary, to maintain 40 to 50 psig in the SW Headers. 2. IF a rupture in a SW Header has occurred, THEN refer to AOP-022. 3. IF an increase in SW temperature has caused SW cooling valves to throttle open, THEN locally throttle SW-739 AND SW-740 as necessary to maintain SW pressure 40 psig to 50 psig. DEVICE/SETPOINTS

1. PSL-1684 I 40 psig POSSIBLE PLANT EFFECTS 1. Loss of Service Water 2. Overheat of CCW 3. Possible entry into TECH SPEC LCO REFERENCES

1. ITS LCO 3.7.7 2. AOP-022, Loss of Service Water 3. CWD B-190628, Sheet 840, cable M 4. Flow Diagram G-190199 I APP-008 Rev.3? Page 50 of 51

24. HLC-06 NRC Replacement Exam Given the following:

-The plant is at 1 00% RTP. -An approved radioactive liquid waste release is in progress. Which ONE (1) of the following correctly describes plant response if detector high voltage is lost to RMS Monitor R-18, Liquid Effluent Waste Disposal? A'I APP-036-E7 RAD MONITOR TROUBLE" actuates and the release automatically terminates. B. APP-036-E7 RAD MONITOR TROUBLE" actuates and the release continues. C. Local Waste Disposal Panel alarm actuates and the release automatically terminates. D. Local Waste Disposal Panel alarm actuates and the release continues. r::.:---------------------.-----

25. Given the following:

/,'-"" -A Reactor Trip has occurred. -30 seconds following the trip, Service Water pressure is 31 psig and slowly decreasing. -40 seconds following the decrease of Service Water pressure, Safety Injection actuates. Which ONE (1) of the following describes the status of the Service Water System? Service Water to the Turbine Building will isolate ... A. 60 seconds after the Reactor Trip. B'I 60 seconds after Service Water pressure decreases below 31 psig. C. immediately upon Service Water pressure decreasing below 31 psig. D. immediately upon the Safety Injection actuation. '----.... _ ... -....... _ ..... - --. ---.... Thursday, October 04,20073:53:51 PM 15 HLC-08 NRC Written Exam 15. Given the following: -The plant is operating at 12% RTP. -A line break in the Instrument Air header has occurred. -IA header pressure is 82 PSIG and decreasing slowly. -The crew has implemented AOP-017, LOSS OF INSTRUMENT AIR. Which ONE (1) of the following will require the crew to manually trip the reactor and enter PATH-1 lAW AOP-017? A. Letdown isolates. B. FCV-1740, IA DRYER BYPASS Valve has OPENED. C. RCP seal injection flows increase to 15 GPM/Pump. D. IA header pressure decreases to 58 PSIG. 15 HLC-08 NRC Written Exam 15.000065 AA2.05 OOllLOSS OF INSTR AIRllI1I3.4/4.1/ROILOWINIAINEW -2008/AOP-017-004 Given the following: -The plant is operating at 12% RTP. -A line break in the Instrument Air header has occurred. -IA header pressure is 82 PSIG and decreasing slowly. -The crew has implemented AOP-017, LOSS OF INSTRUMENT AIR. Which ONE (1) of the following will require the crew to manually trip the reactor and enter PATH-1 lAW AOP-017? A. Letdown isolates. B. FCV-1740, IA DRYER BYPASS Valve has OPENED. C. RCP seal injection flows increase to 15 GPM/Pump. IA header pressure decreases to 58 PSIG. The correct answer is D. A: Incorrect -AOP-017, Section A has operators check if letdown has isolated, but does not direct a Rx trip. B: Incorrect -FCV-1740 OPENS automatically at 80 PSIG, AOP-017 directs the operator to check it OPEN, but the Reactor is not tripped until air header pressure reaches 60 PSIG. C: Incorrect -AOP-017, Section A is for MODE 1/2 operation, Step directs operators to check Seal Injection flow between 8-13 GPM, but allows expanded range of 6-20 GPM. 0: Correct -AOP-017 directs an immediate Rx Trip if IA decreases to 60 PSIG. Exam Question Number: 15

Reference:

AOP-017, Pages 4,5,11-12. KA Statement: Ability to determine and interpret the following as they apply to the Loss of Instrument Air: When to commence plant shutdown if instrument air pressure is decreasing. History: Tuesday, June 17,20081:21:04 PM 18 Rev. 35 AOP-017 LOSS OF INSTRUMENT AIR INSTRUCTIONS

1. Check Plant Status -MODE 1 OR MODE 2 Page 4 of RESPONSE NOT OBTAINED Go To Step 4. (

• 2:) Check IA Header Pressure -LESS -? IF IA pressure decreases to less THAN 60 PSIG than 60 psig. THEN Go To Step 3. 3. Perform The Following:

a. Trip the Reactor b. Go To PATH-I. while continuing with this procedure

4. Verify Instrument Air Compressor D -RUNNING 5. Verify The Primary Air Compressor

-RUNNING

• 6. Check IA Header Pressure -LESS THAN 80 PSIG Go To Step 4. IF IA pressure decreases to less than 80 pSig. THEN observe NOTE prior to Steps 7 and 8 and perform Steps 7 and 8. Observe the NOTE Prior To Step 9 and Go To Step 9. 61 Rev. 35 AOP-017 LOSS OF INSTRUMENT AIR Page 5 of INSTRUCTIONS RESPONSE NOT OBTAINED IA-3821 is located on IA Dryer D. 7. Dispatch Operator(s)

To Perform The Following:

a. Verify Station Air Compressor

-IN SERVICE WITH DISCHARGE VALVE OPEN b. Verify the following SA TO IA CROSS CONNECT BYPASS FILTER ISOLATION Valves -OPEN:

• SA-220
• SA-221 c. Verify IA-18. AIR DRYER "A" & "B" BYPASS -OPEN d. Verify the following Compressors

-RUNNING

• STATION AIR COMP
• INST AIR COMP A
• INST AIR COMP B e. Check FCV-1740.

AIR DRYER HIGH DP FLOW CONTROL Valve -OPEN f. Open IA-3821. INSTRUMENT AIR DRYER "D" BYPASS a. Go To Step 7.c. b. Open SA-5. STATION AIR TO INST AIR CROSS CONNECT. e. Open IA-3665. AIR DRYER "A" & "B" BYPASS. 61 Rev. 35 AOP-017 LOSS OF INSTRUMENT AIR Page 11 INSTRUCTIONS RESPONSE NOT OBTAINED SECTION A Modes 1 AND 2 (Page 1 of 7)

• 1. Determine If IA Capacity Has Been Restored As Follows: a. Check IA Header pressure:
• GREATER THAN 85 PSIG
• STABLE OR INCREASING

b. Go To Attachment

4. Restoration From Loss Of Instrument Air Check Any SIG Level Control -ADVERSELY AFFECTED BY LOSS OF IA a. IF IA capacity is restored.

THEN Go To Step 1.b. Go To Step 2. IF any SIG level control is affected. THEN Go To Step 3. of 61 Observe NOTE prior to Step 4 and Go To Step 4. 3. Perform The Following:

a. Trip the Reactor b. Go To PATH-I. while continuing with this procedure

c. Go To Section B. Hot Shutdown (Without RHR In Service) Use of the RWST for RCS Makeup will add negative reactivity.

• 4. Check VCT Level -LESS THAN 12.5 INCHES IF VCT level decreases to less than 12.5 inches. THEN Go To Step 5. Go To Step 7.

Rev. AOP-017 LOSS OF INSTRUMENT AIR INSTRUCTIONS SECTION A Modes 1 AND 2 (Page 2 of n 5. Align Charging Pump Suction From The RWST As Follows: Page RESPONSE NOT OBTAINED 35 12 of 61 a. From the RTGB. verify LCV-115B. EMERG MU TO CHG SUCT -OPEN a. Open CVC-358. RWST TO CHARGING PUMP SUCTION. prior to continuing.

b. Verify LCV-115C.

VCT OUTLET -CLOSED 6. Perform The Following:

a. Trip the Reactor b. Go To PATH-1. while continuing with this procedure

c. Go To Section B. Hot Shutdown (Without RHR In Service)

Any Of The Following -ADVERSELY AFFECTED: 1J1S+ft I

• Letdown flow indicated on FI-150. LOW PRESS LTDN FLOW
• Letdown pressure indicated on PI-145. LOW PRESS LTDN PRESS
• Letdown temperature indicated on TI-140. REGEN HX LTDN OUTLET TEMP 8. Verify LCV-460 A & B. LTDN LINE STOP Control Switch -CLOSED ----------" ( lF is affected.

THEN Step 8. Go To Step 9. "---" HLC-08 NRC Written Exam 16. Given the following: -The Reactor has tripped due to a LOCA. -The crew has entered PA TH-1. Which ONE (1) of the following would justify entry into EPP-20, LOCA OUTSIDE CONT AINMENT? A. Auxiliary Building Sump level rapidly increasing. B. Abnormal radiation in the Auxiliary building. C. RWST level at 27% with the CV Sump level at 350 inches. D. Decreasing CV Pressure with a stable or increasing RCS pressure. 16 HLC-08 NRC Written Exam 16. WE04 EK3.1 OOllLOCA OUTSIDE CONTAIN/1/1/3.4/4.1IROILOW/NINNEW -2008IEPP-20-002 Given the following: -The Reactor has tripped due to a LOCA. -The crew has entered PATH-1. Which ONE (1) of the following would justify entry into EPP-20, LOCA OUTSIDE CONTAINMENT? A. Auxiliary Building Sump level rapidly increasing. B!o" Abnormal radiation in the Auxiliary building. C. RWST level at 27% with the CV Sump level at 350 inches. D. Decreasing CV Pressure with a stable or increasing RCS pressure. The correct answer is B. A: Incorrect -Many different systems could cause a sump level increase. Rising sump level is NOT an entry condition for EPP-20. B: Correct -The ONLY entry condition to enter EPP-20 is abnormal radiation in the Aux building that is due to a loss of inventory from the RCS. (LOCA) , C: Incorrect -27% is the entry condition for EPP-9, TRANSFER TO COLD LEG RECIRCULATION. 350 inches is below the pOint where Supplement 0 would be used to initiate cold leg recirculation. 0: Incorrect -Increasing RCS pressure is the correct parameter used to ensure that the RCS leak outside Containment is isolated. Exam Question Number: 16

Reference:

EPP-20, Page 3; EPP-17, Page 36. KA Statement: Knowledge of the reasons for the following responses as they apply to the (LOCA Outside Containment): Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics. History: New -Written for HLC-08 NRC Exam. Wednesday, June 18,200810:15:38 AM 19 Rev. EPP-20 LOCA OUTSIDE CONTAINMENT Page Purpose and Entry Conditions (Page 1 of 1) 1. PURPOSE This procedure provides actions to identify and isolate a LOCA outside Containment.

2. ENTRY CONDITIONS 8 3 Path-I, when there is abnormal radiation in the auxiliary building due to a loss of RCS inventory outside Containment.

-END -of 9 EPP-l? SGTR WITH LOSS OF REACTOR COOLANT: SUBCOOLED RECOVERY ATTACHMENT 1 CONTAINMENT SUMP LEVEL VS. RWST LEVEL Page 1 of 1 INDICATED RWST LEVEL (%) 100 .......... ..... ........... ........... 90 --...... i"""--... .......... EXPECTED REGION I ......1'--... ........... .......... I 80 70 I SAFE I

• 60 ...... .........

.... 50 ... -.......... ...... 40 i .... \

• TRANSITION TO EPP-18 I , 30 I , 20 I UNSAFE II 10 I , \ \ \ o , o 48 96 144 192 240 288 336 384 INDICATED CONTAINMENT WATER LEVEL (IN.) Rev. 18 Page 36 of 40 432 480 528 HLC-08 NRC Written Exam 17. Given the following:

-The plant has experienced a Reactor Trip and LOCA, followed by a Loss of Emergency Coolant Recirculation. -RHR is NOT available for core cooling. -The crew has implemented EPP-15, LOSS OF EMERGENCY RECIRCULATION, and is ready to perform Step 18, "Initiate RCS Cooldown to Cold Shutdown". EPP-15, Step 18 directs the crew to cooldown the RCS at _(a)_, using _(b)_ S/Gs. A. (a) maximum rate (b) only intact B. (a) less than 100 of/hour (b) any intact preferred, but faulted if necessary C. (a) maximum rate (b) any intact preferred, but faulted if necessary D. (a) less than 100 of/hour (b) only intact 17 HLC-08 NRC Written Exam 17. WEll EK1.2 OOlILOSS OF EMER RECIRC/1I1I3.6/4.1IROILOW/NINNEW -2008/EPP-lS-003 Given the following: _ The plant has experienced a Reactor Trip and LOCA, followed by a Loss of Emergency Coolant Recirculation. -RHR is NOT available for core cooling. -The crew has implemented EPP-15, LOSS OF EMERGENCY RECIRCULATION, and is ready to perform Step 18, "Initiate RCS Cooldown to Cold Shutdown". EPP-15, Step 18 directs the crew to cooldown the RCS at _(a)_, using _(b)_ S/Gs. A. (a) maximum rate (b) only intact B:I (a) less than 100 °Flhour (b) any intact preferred, but faulted if necessary C. (a) maximum rate (b) any intact preferred, but faulted if necessary D. (a) less than 100 °Flhour (b) only intact The correct answer is B. A: Incorrect -Cooldown at the maximum rate is used to get the RCS at target temperature for a SGTR per FRP-C.1, Step 21. B: Correct -Step 18 directs the crew to "Maintain cooldown rate in RCS cold legs less than 100 of in the last 60 minutes." and to "Check at least one SIG available for cooldown". Step 18 RNO specifies that if RHR is NOT available, then a faulted SIG may be used. c: Incorrect -Rate is incorrect, but SIG to use is correct. D: Incorrect -Rate is correct, but may use a faulted SIG if NO intact SIG is available. Tuesday, June 17, 2008 1 :21 :04 PM 20 HLC-08 NRC Written Exam Exam Question Number: 17

Reference:

EPP-15; EPP-15-BD; FRP-C.1, Page 10. KA Statement: Knowledge of the operational implications of the following concepts as they apply to the (Loss of Emergency Coolant Recirculation): Normal, abnormal and emergency operating procedures associated with (Loss of Emergency Coolant Recirculation). History: New -Written for HLC-08 NRC Exam. Tuesday, June 17,20081:21:04 PM 21 ,"-Rev. 17 EPP-15 LOSS OF EMERGENCY COOLANT RECIRCULATION Page 9 of INSTRUCTIONS RESPONSE NOT OBTAINED

• A differential pressure of 210 psid across the RCP number 1 seals is necessary for continued RCP operation.
• RCS cooldown should be completed as quickly as possible since the RCS may continue to depressurize to a value that may not support differential pressure across the RCP number 1 seals. 18. Initiate RCS Cooldown To Cold Shutdown As Follows:

cooldown rate in RCS cold legs -LESS THAN 100°F IN THE LAST 60 MINUTE b. Maintain RCS temperature and pressure -WITHIN LIMITS OF CURVE 3.4, REACTOR COOLANT SYSTEM PRESSURE -TEMPERATURE LIMITATIONS FOR COOLDOWN c. Check RHR System -ALIGNED FOR CORE COOLING d. Cooldown using RHR System e. Go To Step 24 intact SIGs -AT LEAST ONE AVAILABLE FOR RCS COOLDOWN g. Check steam dump to Condenser -AVAILABLE

h. Dump steam to Condenser from SIGs 19. Check RCS Hot Leg Temperatures

-LESS THAN 543°F c. Go To Step 18.f. RHR System unavailable. THEN use a faulted SIG for RCS cooldown.

g. Dump steam from SIGs using STEAM LINE PORVs. Go To Step 19. WHEN RCS hot leg temperatures less than 543°F, THEN Go To Step 20. 34 Rev. 17 FRP-C.1 RESPONSE TO INADEQUATE CORE COOLING Page 10 of 28 INSTRUCTIONS RESPONSE NOT OBTAINED

• Partial uncovery of SIG tubes is acceptable in the following steps due to steaming faster than feeding.
• After the Low Steamline Pressure SI Signal is blocked, main steamline isolation will occur if the high steam flow rate setpoint is exceeded.
• 21. Depressurize All Intact SIGs To 140 PSIG As Follows: a. Check Steam Dump to Condenser

-AVAILABLE

b. Dump steam to maximum rate c. Check RCS Hot Leg Temperatures

-LESS THAN 543°F d. Defeat Low Tavg Safety Injection Signal as follows: 1) Momentarily place SAFETY INJECTION T-AVG Selector Switch to BLOCK position 2) Verify LO TEMP SAFETY INJECTION BLOCKED status light -ILLUMINATED

e. Check SIG pressures

-LESS THAN 140 PSIG a. Dump steam at maximum rate using STEAM LINE PORVs. Go To Step 21.c. c. WHEN RCS hot leg temperatures less than 543°F, THEN perform Step 21.d. Go To Step 21. e. e. IF SIG pressure is decreasing, THEN observe NOTE prior to Step 19 and Go To Step 19. IF SIG pressure is increasing, THEN Go To Step 28. (CONTINUED NEXT PAGE) HLC-08 NRC Written Exam 18. Which ONE (1) of the following is the reason for placing the Governor valves on the Limiter while at 100% RTP? A. Reduce the effects of turbine overspeed in the event of a load rejection or Generator output breaker trip. B. Prevent Generator overload resulting from Turbine Impulse channel failures and malfunctions. C. Prevent Generator VAR fluctuations induced from Impulse stage pressure fluctuations. D. Minimize Generator load fluctuations to prevent Reactor power from exceeding license limit. 18 HLC-08 NRC Written Exam 18.000077 AK2.06 001lGEN VOLT & GRID DIST/1I1I3.9/4.0/ROILOWININNEW -20081EHC-006 Which ONE (1) of the following is the reason for placing the Governor valves on the Limiter while at 100% RTP? A. Reduce the effects of turbine overspeed in the event of a load rejection or Generator output breaker trip. B. Prevent Generator overload resulting from Turbine Impulse channel failures and malfunctions. C. Prevent Generator VAR fluctuations induced from Impulse stage pressure fluctuations. Minimize Generator load fluctuations to prevent Reactor power from exceeding license limit. The correct answer is D. A: Incorrect -Overspeed and / or Load Rejection will NOT be affected by the position of the Limiter. B: Incorrect -Impulse channel effects are prevented by placing control to "IMP OUT" C: Incorrect -VAR fluctuations can occur regardless of valve position limit. D: Correct -Placing the Governor Valves on the Limiter after reaching steady state load prevents the Turbine Generator's internal Speed Control from trying to maintain 60 Hz if Grid frequency drops. Exam Question Number: 18

Reference:

AOP-026 BD, Page 6. KA Statement: Knowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following: Reactor power. History: New -Written for HLC-08 NRC exam. Tuesday, June 17, 2008 1 :21 :05 PM 22 Step 6 7 \ 10 11 12 13 14 15 BASIS DOCUMENT, GRID INSTABILITY Description This step checks if the Turbine is on-line. If not on-line the RNO bypasses the steps addressing Turbine response. This continuous action step checks if Turbine load has increased due to the decrease in system frequency. If load has increased, the next few steps provide the operator with instructions to limit the load increase and stabilize the plant following the load change. If load has not increased the RNO bypasses these steps. . This step checks if Reactor power is less than or equal to 100%. If power has increased to greater than 100%, the RNO directs the reducing load, with the Valve Position Limiter, to maintain Reactor power less than 100%. The limiter must be used in this condition to set a fixed valve position. The governor speed.-> c;ontrol s'ystem will attempt to maintain RPM by opening the valves, This step directs the operator to limit the Turbine load increase based on current operating equipment ysing the Valve Position limiter. The governor speed control system will attempt to maintain RPM by opening the control valves. The step is dependent on the number of pumps etc operating. If feedwater flow is only sufficient to support 50%, then power must be limited to less than 50%. Checking S/G levels in the normal operating range ensures the plant is stabilizing from the load change and that the automatic level control systems are operating properly. This step checks Tavg trending to Tref. If not, some means of reactivity must be inserted to increase Tavg to Tref. Withdrawing Control rods in manual, or diluting the RCS will increase Tavg. Checking Pressurizer pressure in its normal operating range ensures the plant is stabilizing from the load change and that the automatic pressure control system is functioning properly. Checking Pressurizer level at or trending to program level ensures the plant is stabilizing from the load change and that the automatic level control system is operating properly. This continuous action step checks if a Turbine load reduction has occurred or is in progress. When spinning reserve starts picking up load or customers are being shed to reduce the load on the grid, the frequency may increase above 60 Hz. Being above 60 Hz means being greater than 1800 RPM. The governor control system will close the control valves attempting to maintain speed. This results in a reduction of load. If no load reduction has occurred the RNO bypasses the steps associated with a load reduction. If a load reduction has occurred, this step checks how much load has been reduced to determine what actions need to be taken to stabilize the plant. If the reduction exceeds 100 MWe, the RNO directs the operator to AOP-015, Turbine Runback or Secondary Load Rejection while continuing with this procedure. I AOP-026-BD Rev. 9 Page 6 of 11 I HLC-08 NRC Written Exam 19. Given the following: -The plant is at 45% power, with power escalation to 100% RTP in progress. -Control Bank "0" rods are at 185 steps on the step counters. -Rod Control is in AUTOMATIC. -The RO notes ALL 5 Control Bank "0" rods' IRPI indicators are increasing. -T AVG is 1.5 of higher than T REF' The described conditions indicate ... A. a voltage or frequency problem on Instrument Bus 7A, is causing erroneous IRPI indications. B. a Control Bank "0" Signal Conditioning Module fault is causing erroneous IRPI indications. C. a Pulse to Analog Converter fault is causing erroneous step counter indications. D. an Uncontrolled Rod Withdrawal is occurring. 19 HLC-08 NRC Written Exam 19.000001 AK2.08 OOllCONT. ROD WITHDRAWALllI1I3.1I3.0IROIHIGHlNIAJNEW -2008/AOP-00I-005 Given the following: -The plant is at 45% power, with power escalation to 100% RTP in progress. -Control Bank "0" rods are at 185 steps on the step counters. -Rod Control is in AUTOMATIC. -The RO notes ALL 5 Control Bank "0" rods' IRPI indicators are increasing. -T AVG is 1.5 of higher than T REF, The described conditions indicate ... A. a voltage or frequency problem on Instrument Bus 7A, is causing erroneous IRPI indications. B. a Control Bank "0" Signal Conditioning Module fault is causing erroneous IRPI indications. C. a Pulse to Analog Converter fault is causing erroneous step counter indications. D'r' an Uncontrolled Rod Withdrawal is occurring. The correct answer is O. A: Incorrect -Instrument Bus 7A supplies power to ALL IRPls. Only CB "0" IRPls are changing. B: Incorrect -There is NO single SCM for a rod bank or group. Each IRPI has its own SCM. C: Incorrect -P to A Converter would affect the Rod Bottom Bypass Bistable, but NOT the Step Counters. 0: Correct -RNP has NO automatic Rod Withdrawal, Rods should NOT be moving. Exam Question Number: 19

Reference:

SD-009, IRPI, Figure 22. KA Statement: Knowledge of the interrelations between the Continuous Rod Withdrawal and the following: Individual rod display lights and indications. History: New -Written for HLC-08 NRC exam. Tuesday, June 17,20081 :21 :05 PM 23 -A-Tovg Tavg -C-Tavg TA VG CONTROL BLOCK DIAGRAM RDCNT-FIGURE-22 Nuclear Power (QN) I L .. Powllr Mismatch Rate Compensation Turblnll Load Repreaentlld by Impluse Chamber Prllssurll (Qro) Non-Llnllar Gain K,

• T ref Program '---------

..... -1 b:L U 100" Mlldlan Signal Selector r--Variable Gain h--===-J Trllt Compen 1-----satlon -I Lead/l..ag I--L-Control T avg Compensation Filter I--,...-t--Summing Unit Pressurizer Pressure Input (defeated) INFORMATION USE ONLY Rod Splled Program h-iyfC Direction Bistabills Rods In Rods Out i Rod Control Logic Cabinet HLC-08 NRC Written Exam 20. Given the following: -The plant has experienced a Reactorffurbine Trip from 100% RTP*. -The crew is performing EPP-4, REACTOR TRIP RESPONSE. -Rod Bottom Light for Control Rod M6 is NOT lit. -IRPI shows Control Rod M6 at 225 steps. Which ONE (1) of the following actions, if any, must be performed? A. Adequate shutdown margin exists. Perform actions lAW AOP-001, MALFUNCTION OF REACTOR CONTROL SYSTEM. B. Adequate shutdown margin does NOT exist. Borate for worth of most reactive rod. C. Adequate shutdown margin does NOT exist. Initiate boration to Cold Shutdown requirements. D. Adequate shutdown margin exists. No action required. 20 HLC-08 NRC Written Exam 20. 000005 AK3.01 OOl/INOPERABLE/STUCK ROD/l/2/4.0/4.3IROIHIGHIN/NFARLEY -2001IEPP-4-004 Given the following: -The plant has experienced a Reactorrrurbine Trip from 100% RTP. -The crew is performing EPP-4, REACTOR TRIP RESPONSE. -Rod Bottom Light for Control Rod M6 is NOT lit. -IRPI shows Control Rod M6 at 225 steps. Which ONE (1) of the following actions, if any, must be performed? A. Adequate shutdown margin exists. Perform actions lAW AOP-001, MALFUNCTION OF REACTOR CONTROL SYSTEM. B. Adequate shutdown margin does NOTexist. Borate for worth of most reactive rod. C. Adequate shutdown margin does NOT exist. Initiate boration to Cold Shutdown requirements. Adequate shutdown margin exists. No action required. The correct answer is O. A: Incorrect -AOP-001 is NOT a concurrent AOP and EPP-4 would initiate ALL required actions. B: Incorrect -Adequate SOM does exist, accident analysis assumes the most reactive rod is stuck out, NO action is requred. C: Incorrect -Adequate SOM does exist, this is the required action for TWO or more rods stuck out. D: Correct -Accident Analysis supports adequate shutdown margin with one rod stuck out. Exam Question Number: 20

Reference:

EPP-4, Pages 10 and 11; EPP-4 BO, Pages 8 and 9. KA Statement: Knowledge of the reasons for the following responses as they apply to the Inoperable / Stuck Control Rod: Boration and emergency boration in the event of a stuck rod during trip or normal evolutions. History: Modified answer to meet RNP procedures, changed answer from boration required (Farley requirement) to no boration required. Tuesday, June 17, 2008 1 :21 :05 PM 24 Rev. 22 EPP-4 REACTOR TRIP RESPONSE Page 10 of 28 INSTRUCTIONS RESPONSE NOT OBTAINED ************************************************************************** CAUTION The boration pathway through FCV-114B does NOT have heat trace. Use of this pathway without flush water could result in blockage of the pathway. ************************************************************************** @) Check All Control Rods -FULLY IF only one Control Rod is stuck INSERTED To Step 14. IF two or more Control Rods are stuck out, perform the following:

a. Verify at least one Charging Pump is RUNNING. b. Borate to cold shutdown boron concentration using one of the following:

• Blender to Charging Pump suction: 1) Open FCV-113A, BA TO BLENDER. 2) Open FCV-113B, BLENDED MU TO CHG SUCT. 3) Start Boric Acid Pump aligned for blend.
• RWST to Charging Pump suction: (CONTINUED NEXT PAGE) 1) Open LCV-115B, EMERG MU TO CHG SUCT, OR locally open CVC-358, RWST TO CHARGING PUMP SUCTION. 2) Close LCV-115C, VCT OUTLET.

Rev. 22 EPP-4 REACTOR TRIP RESPONSE INSTRUCTIONS

13. (CONTINUED)

14. Check PZR Level -LESS THAN 14% Page 11 RESPONSE NOT OBTAINED

• Blender to VCT: 1) Open FCV-113A, BA TO BLENDER. of 28 2) Open FCV-114B, BLENDED MU TO VCT. 3) Start Boric Acid Pump aligned for blend.
• Emergency boration:

1) Open MOV-350, BA TO CHARGING PMP SUCT. 2) Start Boric Acid Pump aligned for blend. 3) Verify boric acid flow on FI-110. c. Open CVC-310B, LOOP 2 COLD LEG CHG. IF CVC-310B will NOT open, THEN open CVC-310A, LOOP 1 HOT LEG CHG. d. Verify charging flow on FI-122A. Go To Step 16.

RNP WOG BASIS/DIFFERENCES STEP STEP 12 2 WOG BASIS C13 N/A 13 3 I EPP-4-BD PURPOSE: To ensure the proper feedwater alignment following a reactor trip BASIS: Verification of main feedwater isolation after RCS average temperature reaches the setpoint is necessary to prevent an uncontrolled RCS cooldown from excessive feeding of the steam generators. Verifying feed flow to the steam generators ensures a secondary heat sink for decay heat removal. The feedwater source may be from either the AFW pumps or main FW on the bypass lines. RNP DIFFERENCES/REASONS The RNP step is marked as a Continuous Action step, while the ERG does not indicate Continuous Action. The ERG step is worded as a Continuous Action, but not indicated. This corrects an error in the ERG. Sub-step c of the RNP step establishes flow to maintain level in the normal control band rather than a minimum flow from Main or Auxiliary Feedwater. This will accommodate trips from low or no power level in which S/G shrink would be insufficient to drive level low. If feedwater is established at the minimum flow specified by the ERG in these cases, S/G overfeed will result. Establishing a control band allows the operator to vary the amount needed The intent to establish required flow is maintained in the RNP step. SSD DETERMINATION This is an SSD per criterion 8 and 11. WOG BASIS N/A, this step is not in the WOG ERG. RNP DIFFERENCES/REASONS This caution is not contained in the ERG. The caution is provided as a warning to remind the operator of the fact that there is no heat tracing in the Boric Acid line to the top of the VCT and the consequences of boration through this line without the ability to flush. One of the potential pathways for boration is via the top of the VCT. This could be a problem if a LOOP has occurred since the PW Pumps are powered from a non-vital bus. SSD DETERMINATION This is an SSD per criterion

10. WOG BASIS PURPOSE: To ensure all control rods are inserted for adequate shutdown margin BASIS: .. _----, .. _--"--:> A subcritical core is verified if all rods are at the bottom according to the rod bottom lights and the rod position indicators.

If these indications reveal that one rod is not inserted, no immediate action is re uired since the core is desi ned for ade uate shutdown margm--' Wit one rod stuck out. However, if more than one rod fails to insert fullY, the shut own reactivity margin must be made up tiirough emergency boration to account for the reactivity worth of the stuck rods. Rev. 22 Page 8 of 19/ RNP WOG BASIS/DIFFERENCES STEP STEP 14-17 4 N23 N/A I EPP-4-BD RNP DIFFERENCES/REASONS The steps accomplish the same objective. The EPP provides detailed instructions on emergency boration and also provides alternative actions to get boron to the core. The ERG directs the operator to emergency borate "x" number of ppm for each control rod not fully inserted. The EPP directs the operator to emergency borate to.£Qld This assures reactor shutdown regardless of the number of control rods not fully inserted. This guidance was provided by Reactor Engineering personnel to allow for a conservative value for multiple rods which causes varying rod worth, dependent on the location, number, and core life. SSD DETERMINATION This is an SSD per criterion 5 and 10. WOG BASIS PURPOSE: To ensure normal post-trip pressurizer level response BASIS: Following the reactor trip, pressurizer level is expected to decrease to the no-load value due to shrinkage of the RCS from the at-power, programmed temperature to no-load temperature. The pressurizer level control system should then stabilize level at the load value. If level decreases below the no-load value, the operator should try to reestablish no-load level by operation of the charging system. If level drops below the letdown isolation setpoint, the operator should also verify letdown isolation and restore level prior to reestablishing letdown. Level should then be controlled at the no-load value. RNP DIFFERENCES/REASONS The ERG step has been split into 4 steps for the RNP procedure due to the large amount of additional information required. The decision criteria was changed from "GREATER THAN" to "LESS THAN" to remove the NOTs from RNO column. Letdown is placed in the CLOSE position to prevent the letdown valves from automatically opening and lifting the letdown relief valve. The EPP contains supplemental information on the operation of PZR heaters that is particular to plant design. When the heaters are de-energized due to letdown isolation, re-energizing the heaters after level is regained requires the control switch to be placed in the OFF position and returned to the ON position. With RCP B running and RCP C not running, and pressurizer level less than 30%, PZR spray flow would be insufficient. The operator is directed to control level between 30% and 40%, if this condition exists, and transition to Step 16. This step will take care of all permutations of RCPs running. ERG Step 4 requires maintaining PZR level at 22%. RNP Step 15 was inserted to control level higher if RCP C not running, this step was added to address the condition for C RCP running. SSD DETERMINATION This is an SSD per criterion 4 and 10. WOG BASIS N/A, this step is not in the WOG ERG. RNP DIFFERENCES/REASONS Auxiliary Spray will be used in the subsequent step. The note reminds the operator that Supplement K is available for use if desired. Rev. 22 Page 9 of 19/ Alarm DDl, "RCP SEAL INJ FLOW LO," annunciated a short time after LT-459 failed low. Which ONE of the following describes the reason for the above alarm? A. The isolation of letdown caused charging flow to decrease and the seal injection header flow diminished due to the increasing pressure on the charging header. B. The loss of the pressurizer heaters caused charging flow to decrease and the seal injection header flow diminished due to the increasing pressure on the charging header. C. As charging flow increased, the DIP across the seal injection filter rose and the seal injection header flow diminished due to the increased resistance to flow. D. As charging flow increased, the seal injection header flow diminished due to the decreasing pressure on the charging header. A -Incorrect, Increased charging header pressure would cause the seal injection flow to increase. B -Incorrect, Increased charging header pressure would cause the seal injection flow to increase. C -Incorrect, Increased DIP across the seal injection filter under the stated conditions could only be caused by increased flow through the filter. D -Correct, Increased charging flow robs the seal injection line of flow. Source: Farley Bank Question # 052101D14044 Answer: D 12. 005AK1.05 1 Unit 1 has experienced a reactor/turbine trip from full power. -----."----------- 10\ N<<'/-'

---

While performing FNP-I-ESP-O.l, "REACTOR TRIP RESPONSE," it is observed that the 'Rod Bottom' light is NOT lit for control rod 'M6' in Control Bank 'B'. DRPI shows control rod 'M6' at 228 steps. Which ONE of the following actions must be performed in accordance with FNP-I-ESP-O.l, "REACTOR TRIP RESPONSE," in response to this condition?

A. No action is required. B. An emergency boration of at least 2500 gallons is required. C. An emergency boration of at least 5000 gallons is required. D. Verify shutdown margin within the limits provided in the COLR. A -Incorrect, Even a single rod stuck in the core, post trip, requires an emergency boration. B -Correct, ESP-0.1 Step 3 RNO Step 3.5 requires at least 2500 gal boration per rod not inserted. C -Incorrect, This value would be for 2 rods not fully inserted. D -Incorrect, This is the TS action for one or more rods untrippable. Source: Summer NRC Exam 2000-301 Answer: B 13. 005EG2.4.4 1 Given the following conditions on Unit 1: -Reactor power is at 73%. -Turbine load is being slowly -Control Bank D rod 'BS' indicates 144 steps. -Control Bank D rod 'K6' indicates 156 steps. -Remaining Control Bank D' rods indicate 16S steps. -Control Bank D is being moved as required to maintain Delta I. -Rod 'BS' and does NOT move when Control Bank D is moved 'out' or 'in'. -Rod 'K6' does move when Control Bank D is moved 'out' or 'in'. Which ONE of the following describes the action required to be taken within one hour? A. Be in Mode 3, Hot Standby. B. Reduce turbine load and insert Control Bank D to 155 steps. C. Trip the reactor immediately and go to EEP-O, "REACTOR TRIP OR SAFETY INJECTION." D. Determine that the shutdown margin is within the limits specified in the COLR. A -Incorrect, Must be in Mode 3 in 6 hours. B -Incorrect, This will place the stuck rod within the 12 step limit but will not correct the K6 rod. C -Incorrect, This is correct action for unexplained rod motion or dropped rod. D -Correct, Rod BS could be considered to be untrippable therefore, TS 3.1.4 requires verification of adequate SDM. TS 3.1.4 also requires the verification of adequate SDM if more than one rod not within alignment limit. Source: New Answer: D 14. 005K1.13 1 Given the following conditions on Unit 2: -The injection phase of a LOCA is in progress. -RHR pump flows indicate 1350 gpm on the discharge of each RHR pump. ) HLC-08 NRC Written Exam 21. Given the following: -The plant is operating at 50% RTP following a 50% load rejection from 100% RTP. -APP-003-C8, PZR PROT HI LEVEL is illuminated. -All Pressurizer Level channels, LT-459A, 460, and 461, indicate 92%. -RTB A & B CLOSED (RED) indication is illuminated. Which ONE (1) of the following is the NEXT procedure the crew should implement? A. AOP-015, SECONDARY LOAD REJECTION. B. AOP-025, RTGB INSTRUMENT FAILURE. C. PATH-1. D. FRP-S.1, RESPONSE TO NUCLEAR POWER GENERATION / ATWS. 21 HLC-08 NRC Written Exam 21. 000028 G2.4.4 OOIIPZR L VL MALFUNCTION/1I2/4.5/4.7IROIHIGHJNINNEW -2008IPATH-I-002 .Given the following: -The plant is operating at 50% RTP following a 50% load rejection from 100% RTP. -APP-003-C8, PZR PROT HI LEVEL is illuminated. -All Pressurizer Level channels, L T-459A, 460, and 461, indicate 92%. -RTB A & B CLOSED (RED) indication is illuminated. Which ONE (1) of the following is the NEXT procedure the crew should implement? A. AOP-015, SECONDARY LOAD REJECTION. B. AOP-025, RTGB INSTRUMENT FAILURE. PATH-1. D. FRP-S.1, RESPONSE TO NUCLEAR POWER GENERATION / ATWS. The correct.answer is C. A: Incorrect -PATH-1 would take priority over AOP-015 due to having high PZR levels. B: Incorrect -No instrument failure is indicated. C: Correct -Pressurizer Levels above 91 % (2/3 while greater than P-7) should have initiated a reactor trip. According to OMM-022, EMERGENCY OPERATING PROCEDURES USER'S GUIDE, Section 8.2.1, "Entry into the EOP Network will be required when the following conditions occur: If at any time a reactor trip or safety injection occurs or is required, the Operator will enter PATH-1." D: Incorrect -PATH-1 is required, FRP-S.1 is NOT a direct entry procedure. Exam Question Number: 21

Reference:

APP-003-C8; OMM-022, Page 10, Step 8.2.1.1. KA Statement: Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures. History: New -Written for HLC-08 NRC exam. Tuesday, June 17,2008 1 :21 :05 PM 25 ALARM PZR PROT HI LEVEL AUTOMATIC ACTIONS 1. Reactor trip (2/3 logic interlocked with P-7) CAUSE 1. Letdown-Charging mismatch 2. Load rejection

3. Channel failure 4. Increasing T avg OBSERVATIONS

1. Pressurizer Level (LI-460, LI-461 and LI-459A) 2. Letdown and Charging flow (FI-150 and FI-122A) 3. Tavg 4. Generator load ACTIONS 1. IF Reactor Trip has occurred, THEN Refer To PATH-1, EOP Network. 2. IF a channel has failed, THEN remove from service using OWP-030. 3. Increase Letdown flow and/or decrease charging flow to reduce level DEVICE/SETPOINTS

1. LC-459A/91%

of span 2. LC-460A / 91 % of span 3. LC-461 /91 % of span POSSIBLE PLANT EFFECTS 1. Reactor Trip REFERENCES

1. PATH-1, EOP Network 2. ITS LCO 3.3.1, 3.4.9 3. AOP-015, Secondary Load Rejection

4. OWP-030, Pressurizer Level Transmitters (PL T) 5. CWD B-190628, Sheet 440, Cable BF I APP-003 Rev. 37 APP-003-C8 Page 30 of 531 8.1 .4 (Continued)

4. Applicability

-The AOPs are generally applicable for all plant conditions from Cold Shutdown to full power. The specific applicability of an AOP is listed on the Purpose and Entry Conditions (PEC) page of the procedure. 8.1.5 APP General Layout 1. The APPs are comprised of procedures containing the instructions for each individual alarm function of an Annunciator Panel. 2. The specific format of the APPs is described in AP-007. 3. The APPs provide instruction specific to a single alarm. These alarms mayor may not be indicative of a plant transient. These alarms are generally indicative of a problem with an individual component.

4. Specific guidance for use of the APPs is given in OMM-001-15.

8.2 General Use of The EOP Network NOTE: In all cases of EOP use, the reactor is assumed to either be shutdown or attempts are in progress to shutdown the reactor. IOMM-022

• If at any time a reactor trip or safety injection occurs or is ___

the Oeerator will enter Path-1 . / .. ------------... -* If at any time a complete loss of power on the AC Emergency Busses takes place, the Operator will enter EPP-1, Loss Of All AC Power. This includes any time during the performance of any other EOP procedure.

• EPP-5 may be entered directly whenever natural circulation is required.
• EPP-21 may be entered directly whenever pressurizer heaters must be energized from the Emergency Busses.
• EPP-25 may be entered directly whenever it is required to energize supplemental plant equipment using the DS Bus. Rev. 28 Page 1 0 of 54/

HLC-08 NRC Written Exam 22. I&C has just completed a surveillance on the high voltage power supply to the Source Range nuclear instruments. The surveillance determined the as-found voltage was 1400 VDC, instead of the normal 1600 VDC. Which ONE (1) of the following describes the effect and reason the lower voltage has on SR N-31 and N-32 instrument response? N-31 and N-32 indication will .... A. NOT be affected because the high voltage only supplies power to the channels' amplifiers and electronic circuitry. B. increase due to a reduction of pulse height discriminations that allows more ionization events to pass through the discriminator circuit. C. increase due to a reduction in gamma compensation, allowing more, lower energy events to pass through the pulse height discriminator circuit. D. decrease because reduced voltage results in fewer ion pairs reaching the detector electrodes due to lower potential applied to the detector. 22 HLC-08 NRC Written Exam 22. 000032 AK1.01 OOllLOSS OF SR NII1I2/2.S/3.1IRO/HIGH/N/AfTURKEY POINT -2001lCOMPON CHAP 2-020 I&C has just completed a surveillance on the high voltage power supply to the Source Range nuclear instruments. The surveillance determined the as-found voltage was 1400 VDC, instead of the normal 1600 VDC. Which ONE (1) of the following describes the effect and reason the lower voltage has on SR N-31 and N-32 instrument response? N-31 and N-32 indication will .... A. NOT be affected because the high voltage only supplies power to the channels' amplifiers and electronic circuitry. B. increase due to a reduction of pulse height discriminations that allows more ionization events to pass through the discriminator circuit. C. increase due to a reduction in gamma compensation, allowing more, lower energy events to pass through the pulse height discriminator circuit. decrease because reduced voltage results in fewer ion pairs reaching the detector electrodes due to lower potential applied to the detector. The correct answer is D. A: Incorrect -Indicated power will decrease. The reduction of 200 VDC is significant. B: Incorrect -Pulse height discriminator circuit has no relation to the High Voltage applied to the detector. C: Incorrect -There is NO gamma compensation circuit (IR Only). D: Correct -The high voltage set at 1600 VDC in the Proportional Region of the detector curve, such that a significant reduction in applied voltage will result in a reduced count rate. Exam Question Number: 22

Reference:

SD-010, NI, Pages 16-17, Figure 7. KA Statement: Knowledge of the operational implications of the following concepts as they apply to Loss of Source Range Nuclear Instrumentation: Effects of voltage changes on performance. History: Modified by changing three distractors and updating to to RNP equipment voltages. Tuesday, June 17, 2008 1 :21 :06 PM 26 SD-OlO NUCLEAR INSTRUMENTATION SYSTEM NIS in the NIS. Region I is the Recombination Region which has the following general characteristics:

• Not all ion-pairs formed are collected
• Increasing the applied voltage results in an increase in the number of ion-pairs collected
• The number of ion-pairs formed is dependent upon the type of radiation (i.e. Neutron and Gamma caused ionization will be different)

The IR inner chamber operates in this region, allowing large changes in compensation with small changes in voltage (0-1 IOV) Region II is the Ionization Region which has the following general characteristics:

• All ion-pairs formed are collected
• Increasing the applied voltage results in no increase in the number of ion-pairs collected
• The number of ion-pairs formed is dependent upon the type of radiation (i.e. Neutron and Gamma caused ionization will be different)

The IR outer chamber, PR, and Channels N5I/N52 detectors operate in this region. Region III is the Proportional Region which has the following general characteristics:

• Applied voltage is sufficiently nigh to cause secondary ionizations (called Gas Amplification)
• Output is proportional to the ionizing event
• The Gas Amplification ranges from a factor of 1 to a factor of 10 4 The SR Detectors operate in this range. _ .. The Nuclear Instrumentation System detectors are the gas filled type. The SR detectors use BF3 as a fill gas, the Intermediate and PR Detectors use N2 as fill gas. The fission chambers are also gas filled. When boron is used to detect neutrons by indirect ionization the following interactions occur: The 2He 4 particle is also called an alpha particle.

The Li +3 and He +2 ions induce ionization of gas atoms. The neutrons detected by excore detectors are leakage neutrons Page 16 of 48 Revision 7 INFORMATION USE ONLY SD-OIO NUCLEAR INSTRUMENTATION SYSTEM The Nuclear Instrumentation System employs eight radial detector locations containing a total of sixteen detectors (two Proportional Counters, two Compensated Ionization Chambers, four dual-detector Uncompensated Ionization Chamber assemblies, and two dual-detector fission chambers) installed around the Reactor in the Primary shield (see Figure 1). 3.1.1 SR Detectors BF3 Proportional Counters (see Figure 8) have a nominal thermal neutron sensitivity of ten counts per neutron per square centimeter-second. voltage Vdc with a maximum of 1900Vdc. ------These counters provide pulse signals to the SR Channels. These detectors are installed on opposite sides of the core at an elevation approximating the quarter core height. High density polyethylene is used as a moderator and insulator inside the housing assembly. The SR detectors are designed such that gas pressure and volume minimize the magnitude of pulse signals caused by gamma radiation. The incident neutrons interact with the boron in the BF3 gas producing large pulses. The pulse amplitude created by the neutron is about 6 times larger than that created by the gamma. The gamma radiation does not interact with the boron to produce a large pulse. The output of these detectors is then fed to a discriminator circuit which will not pass the smaller gamma produced pulses, but will pass the larger pulses produced by neutron ionization. 3.1.2 IR Detectors NIS Compensated Ionization Chambers (see Figure 9) serve as neutron sensors for the IR Channels, and are located above, in the same instrument wells and detector assemblies, as the SR Detectors. These detectors have a nominal thermal neutron sensitivity of 4 x 10-14 amperes per neutron per square centimeter-second. Gamma sensitivity is less than 3 x 10-11 amperes per roentgen per hour when operated uncompensated, and is reduced to approximately 3 x 10-13 amperes/R/hr in compensated operation. The detectors are positioned with their centers at an elevation corresponding to one half of the core height. The IR Detectors are constructed to have two N2 filled chambers, one inside the other. Both the inner and outer surfaces of the outside chamber is coated with Boron-lO. The inner chamber has no coating and operates in the recombination region. Both neutron and gamma reactions occur in the outer chamber while only gamma reactions occur in the inner chamber. The output of the inner chamber is subtracted from the output of the outer chamber which results in a net output caused by neutron radiation only. The Page 17 of 48 Revision 7 INFORMATION USE ONLY I-Z w > w z 12 0 10 10 ....... 10 c W I-U w 8 -' -' 0 10 u I/) c.:: 6 z 10 Q u. 0 4 c.:: 10 w a::I :E ::> z TYPICAL NUMBER OF ION PAIRS COLLECTED PER RADIATION EVENT VERSUS VOLTAGE APPLIED ON A FILLED DETECTOR NI-FIGURE-7 c C 0 += c I .... I jl c c QI .2 0 0 .21 t QI C a. C> E N 0 I I 81 '2 a. 0 .2 .... CI.. QI I I] I "I-C I" '<< II III V I VI I I I I I I I I V1 V2 V3 V4 V5 APPLIED VOLTAGE Curve 1: Radiation event of lower specific ionization. Curve 2: Radiation event of higher specific ionization. nif07 Turkey Point Unit 3 I&C just completed a surveillance on the high voltage power source to the Source Range (SR) nuclear instruments. I&C determined the voltage was 1800 Vdc, (200 Vdc lower than the normal 2000 Vdc). Which ONE of the following describes the effect (and the reason) this lower than normal voltage has on SR nuclear instrument performance. SR indication will... D .... decrease because the reduced voltage in the high voltage power supply results in fewer ion pairs reaching the electrodes due to lower potential applied to the detector. A. ... not be affected because the high voltage only supplies power to the electronic circuitry of the amplifier. B ... .increase because the reduced voltage in the high voltage power supply results in more ion pairs reaching the electrodes due to lower potential applied to the detector. C .... decrease because smaller pulses are generated by the alpha decay ofU235 and gamma interactions. A -Incorrect; Indicated power would decrease. The high voltage is set at 2000 V dc in the plateau of SD004 Figure 15 so that small variations in high voltage will not greatly effect the count rate sensed by the detector. 200 Vdc decrease is not a small variation. B -Incorrect; Indicated would decrease because the fewer ions would be measured due to the lower voltage. C -Incorrect; Pulses generated by the alpha decay are unaffected by the voltage, measurement of those pulses is affected. HLC-08 NRC Written Exam 23. Given the following: -The plant is at 30% RTP, with a power escalation in progress. -Power escalation is limited to 1 %/hr due to fuel leakage limitations. -APP-036-C7, R-24 MONITOR HI has just alarmed. R-24, MAIN STEAM LINE N-16 DETECTOR Which ONE (1) of the following describes the condition and the appropriate AOP entry requirements, if any? A. R-24 information is considered invalid with identified fuel leakage. No further actions are required. B. R-24 information is useful only for trending below 40% power. Observe R-24 trends, and if R-15 OR R-19 confirm evidence of leakage, refer to AOP-035, S/G TUBE LEAKAGE. C. R-24 in alarm is an entry condition, crew must enter AOP-035, S/G TUBE LEAKAGE. D. R-24 information below 40% power is useful only for trending. Adjust R-24 leakage model firmware for power level below 40%. 23 HLC-08 NRC Written Exam 23.000037 AA1.06 00l/SGTRl1/2/3.S/3.9/RO/HIGH/N/A/NEW -200S/AOP-035-002 Given the following: -The plant is at 30% RTP, with a power escalation in progress. -Power escalation is limited to 1 %/hr due to fuel leakage limitations. -APP-036-C7, R-24 MONITOR HI has just alarmed. R-24, MAIN STEAM LINE N-16 DETECTOR Which ONE (1) of the following describes the condition and the appropriate AOP entry requirements, if any? A. R-24 information is considered invalid with identified fuel leakage. No further actions are required. B:' R-24 information is useful only for trending below 40% power. Observe R-24 trends, and if R-15 OR R-19 confirm evidence of leakage, refer to AOP-035, S/G TUBE LEAKAGE. C. R-24 in alarm is an entry condition, crew must enter AOP-035, S/G TUBE LEAKAGE. D. R-24 information below 40% power is useful only for trending. Adjust R-24 leakage model firmware for power level below 40%. The correct answer is B. A: Incorrect -R-24 provides useful information, but only when? 40% power. R-24 is an N-16 monitor, does NOT become invalid due to fuel leakage. B: Correct -Entry into AOP-035 is not justified unless R-15 or R-19 indication confirms primary to secondary leakage, due to the limitations of R-24 when power is less than 40%. C: Incorrect -Confirmation of S/G Tube Leakage is required for AOP entry. This would be confirmed by an increasing trend. D: Incorrect -Action to adjust leakage model firmware is to make R-24 valid below 40% power. This may be performed during extended low power operation. Exam Question Number: 23

Reference:

AOP-035, Page 3; APP-036-C7. KA Statement: Ability to operate and / or monitor the following as they apply to the Steam Generator Tube Leak: Main steam line rad monitor meters. History: New -Written for HLC-08 NRC Exam. Tuesday, June 17,20081:21:06 PM 27 AOP-035 SIG TUBE LEAK I Rev. 19 . Page 3 of 52 Purpose and Entry Conditions (Page 1 of 1) 1. PURPOSE This procedure provides the direction necessary to respond to a Steam Generator tube leak when the EOP network has not been entered.

• An alarm on RR-1 is confirmed if any diverse OR redundant indication shows primary to secondary leakage.
• Alarms on R-15. 19. and R-24 are confirmed in accordance with APP-036. Auxiliary Annunciator.

2. ENTRY CONDITIONS Any of the below conditions:

• Confirmed R-24 leakage greater than 30 gpd.
• Confirmed alarm on radiation monitors R-15 OR R-19.
• Confirmed warning alarm on RR-1. RMS RECORDER.

for radiation monitors R-15 OR R-19.

• Notification by Chemistry Personnel that a PSAL-1. PSAL-2. OR PSAL-3 condition exists for primary to secondary leakage. -END -

ALARM R-24 MONITOR HI *** WILL REFLASH *** AUTOMATIC ACTIONS 1. None CAUSE 1. Primary to Secondary S/G Tube leakage while at power 2. Failure of ERFIS power input (causes monitor to default to 40% and increases sensitivity) OBSERVATIONS

1. R-24 Monitor Chart Display 2. Trends of radiation monitors R-15 and R-19. 3. Power Level, NIS and ERFIS APP-036-C7 Page 1 of2 NOTE: IF power is less than 40%, THEN R-24 information is useful only for trending.

Also, the information is not likely to be accurate on the initial alarm at any power level prior to unit adjustment for the leakage location. IF the cause of the alarm is known to be the movement of radioactive material, such as a loaded spent fuel cask, THEN no further actions are required. ACTIONS 1. OBSERVE monitor trends for leak rate AND evidence of short term spiking. 2. IF short term spiking is evidenced, THEN allow the indicated level to decrease prior to continuing. IF short term spiking is present, THEN the alarm is NOT valid. 4. IF trends from R-15 OR R-19 confirm evidence of primary to secondary leakage, THEN REFER to AOP-035. 5. IF neither R-15, nor R-19, show evidence of primary to secondary leakage, THEN PERFORM OP-504 to monitor for low level primary to secondary leakage. 6. IF OP-504 performance confirms evidence of primary to secondary leakage, THEN REFER to AOP-035. 7. CONTACT E&C Personnel to sample the affected S/G to determine the status of primary to secondary leakage .. 8. IF primary to secondary leakage is confirmed, THEN INITIATE action to adjust the R-24 leakage model firmware for the leakage location.

9. WHEN the R-24 leakage model firmware has been adjusted for the leakage location, THEN PERFORM the following:

1) IF indicated leakage is greater than 5 gpd, THEN MONITOR R-24 using ERFIS AND REFER to OMM-001-13.

2) IF indicated leakage is greater than 30 gpd, THEN REFER to AOP-035. 10. IF the alarm is NOT valid, THEN INITIATE action to determine the reason for the alarm. Troubleshoot and repair channel Investigate area for transient radioactive material I APP-036 Rev. 62 Page 23 of 961 HLC-08 NRC Written Exam 24. Which ONE (1) of the following describes the location and type of indication available outside of the Control Room to monitor the availability of Heat Sink? A. Steam Flow and Feedwater Flow for all 3 S/Gs at the Secondary Control Panel. B. All 3 S/G Levels and Pressures at the Charging Pump Room Panel. C. All 3 S/G Levels and Pressures at the Secondary Control Panel. D. Steam Flow and Feedwater Flow for all 3 S/Gs at the Charging Pump Room Panel. 24 HLC-08 NRC Written Exam 24. 000068 AA2.0S OOllCONTROL ROOM EVAC/1/2/4.2/4.3fROILOW/N/A/NEW

-2008/AOP-004-00S Which ONE (1) of the following describes the location and type of indication available outside of the Control Room to monitor the availability of Heat Sink? A. Steam Flow and Feedwater Flow for all 3 S/Gs at the Secondary Control Panel. B. All 3 S/G Levels and Pressures at the Charging Pump Room Panel. All 3 S/G Levels and Pressures at the Secondary Control Panel. D. Steam Flow and Feedwater Flow for all 3 S/Gs at the Charging Pump Room Panel. The correct answer is C. A: Incorrect -There are NO Steam Flow/Feed Flow indications available outside of the Control Room at the Secondary Control Panel. B: Incorrect -S/G Pressure is NOT available at the Charging Pump Room Panel. C: Correct -WR S/G Levels and S/G PORVs and PICs are available at the Secondary Control Panel. D: Incorrect -There are NO Steam Flow/Feed Flow indications available outside of the Control Room at the Charging Pump Room Panel. Exam Question Number: 24

Reference:

AOP-004, Pages 18 and 26. KA Statement: Ability to determine and interpret the following as they apply to the Control Room Evacuation: Availability of heat sink. History: New -Written for HLC-08 NRC exam. Tuesday. June 17,20081 :21 :06 PM 28 Rev. AOP-004 CONTROL ROOM INACCESSIBILITY Page INSTRUCTIONS RESPONSE NOT OBTAINED CONTINUOUS USE ATTACHMENT 2 TURBINE BUILDING OPERATOR (Page 3 of 6) 9. Transfer S/G PORV Control To The Local Controllers At The Secondary Control Panel As. F 11 f [Iers o ows: ., .... 1(1 CHI a. Place PIC-477. manual thumbwheel. to the closed position by rotating the white thumbwheel in the up direction

b. Place PORV RV-l Switch in the DEFEAT position c. Place the transfer switch on PIC-477 to MAN position d. Place PIC-481. manual thumbwheel.

to the closed position by rotating the white thumbwheel in the up direction

e. Place PORV RV-2 Switch in the DEFEAT position f. Place the transfer switch on PIC-487 to MAN position g. Place ,PIC-4CQ, manual thumbwheel.

to the closed position by rotating the white thumbwheel in the up direction

h. Place PORV RV-3 Switch in the DEFEAT position i. Place the transfer switch on PIC-497 to MAN position 16 18 of 30 AOP-004 Rev. CONTROL ROOM INACCESSIBILITY Page INSTRUCTIONS RESPONSE NOT OBTAINED CONTINUOUS USE ATTACHMENT 3 AUXILIARY FEEDWATER OPERATOR (Page 5 of 8) 16 26 of 30 13. Check MDAFW Pump Status -AT Verify the SDAFW Pump is running LEAST ONE RUNNING as follows: 14. Contact The SSO And Request An Additional Operator With A Radio a. Verify MS-Vl-8A, SG "A" STM SUPPLY TO STM DRIVEN AFW PUMP, is OPEN. b. Verify MS-Vl-8B.

SG "B" STM SUPPLY TO STM DRIVEN AFW PUMP, is OPEN. c. Verify MS-Vl-8C, SG "C" STM SUPPLY TO STM DRIVEN AFW PUMP, is OPEN. d. Verify AFW-V2-14A, SDAFW PUMP FW DISCH TO SG "A", is OPEN. e. Verify AFW-V2-14B, SDAFW PUMP FW DISCH TO SG "B", is OPEN. f. Verify AFW-V2-14C, SDAFW PUMP FW DISCH TO SG "C", is OPEN. g. Throttle SDAFW Pump Discharge Valves To Maintain SIG Levels Between 60% And 68% WR Using The SIG WR Level Indicators Control Panel -h. Go To Step 17. IF an additional operator is NOT available, THEN the Auxiliary Feedwater Operator should periodically monitor SIG level AND perform Steps 15 AND 16. HLC-08 NRC Written Exam 25. Given the following: -A Large Break LOCA has occurred. -Containment pressure peaked at 36 PSIG. -BOTH Containment Spray pumps are operating. -Containment pressure and leak rate from Containment continue to decrease. At which Containment pressure will the leak rate from Containment be Y2 the leak rate of the peak pressure? A. 3 PSIG. B. 6 PSIG. C. 9 PSIG. D. 18 PSIG. 25 HLC-08 NRC Written Exam 25. 000069 AK1.01 OOllLOSS OF CONT INTEG/1/2/2.6/3.1/RO/HIGH/N/A/NEW -200S/COMPON CHAP 2-010 Given the following: -A Large Break LOCA has occurred. -Containment pressure peaked at 36 PSIG. -BOTH Containment Spray pumps are operating. -Containment pressure and leak rate from Containment continue to decrease. At which Containment pressure will the leak rate from Containment be Y2 the leak rate of the peak pressure? A. 3 PSIG. B. 6 PSIG. 9 PSIG. D. 18 PSIG. The correct answer is C. A: Incorrect -3 is the square root of the correct answer. B: Incorrect -6 is the square root of the peak pressure. c: Correct -Break flow is proportional to the square root of Delta P. D: Incorrect -18 is half of peak pressure. Exam Question Number: 25

Reference:

GFES, Detectors section of Components, Equation 2-2. KA Statement: Knowledge of the operational implications of the following concepts as they apply to Loss of Containment Integrity: Effect of pressure on leak rate. History: New -Written for HLC-08 NRC exam. Tuesday, June 17, 2008 1 :21 :07 PM 29

• *OIPoc{V)2 Where: DIP = DifferentiaJPressure

... V = Volumetric flow rate {spfm )

• V :::: K,JD1P Where:
• V = Volumetric flow rate (scfm) K = Constant for th\e restriction DIP = Differential pressure PWR / Components

/ Chapter 2 Head flow meters operate on principle that placing restriction in fluiQ stream causes pressure drop The respltingDP is measured to provide flow rate The Qasicprimciple is that when fluid flows through restri.ctiJlfits velbqity willincreaSce anciit& .pressLJrewill decrease Although fluid's velocity increases, its volumetric flow rate does not change The change in pressure (DIP), however,isproportional to squarE) of volumetric flow rate PWR / Components / Chapter 2 HLC-08 NRC Written Exam 26. Given the following: -A Reactor trip has occurred. The crew is preparing to transition from PATH-1 after checking SI NOT actuated OR required. -All S/G levels are 10-14% Narrow Range. -AFW flow indicates 400 GPM. -S/G "A" and "S" pressures indicate 1 080 PSIG. -S/G "C" pressure indicates 1150 PSIG. Which ONE (1) of the following describes the status of CSF-3, HEAT SINK and the initial crew actions to address the event? A. RED; Check total feedwater flow less than 300 GPM due to operator action, lAW FRP-H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK. B. YELLOW; Isolate feedwater to affected S/G lAW FRP-H.2, RESPONSE TO S/G OVERPRESSU RE. C. YELLOW; Check any S/G level greater than 75% lAW FRP-H.3, RESPONSE TO S/G HIGH LEVEL. D. YELLOW; Check level in affected S/G less than 16% lAW FRP-H.5, RESPONSE TO STEAM GENERATOR LOW LEVEL. 26 HLC-08 NRC Written Exam 26. WEl3 EK2.2 OOllSG OVERPRESSURE/l/2/3.0/3.2IROIHIGH/N/A/NEW -200S/FRP-H.2-002 . Given the following: -A Reactor trip has occurred. The crew is preparing to transition from PATH-1 after checking SI NOT actuated OR required. -All SIG levels are 10-14% Narrow Range. -AFW flow indicates 400 G PM. -SIG "A" and "8" pressures indicate 1080 PSIG. -SIG "c" pressure indicates 1150 PSIG. Which ONE (1) of the following describes the status of CSF-3, HEAT SINK and the initial crew actions to address the event? A. RED; Check total feedwater flow less than 300 GPM due to operator action, lAW FRP-H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK. YELLOW; Isolate feedwater to affected SIG lAW FRP-H.2, RESPONSE TO S/G . OVERPRESSURE. C. YELLOW; Check any S/G level greater than 75% lAW FRP-H.3, RESPONSE TO SIG HIGH LEVEL. D. YELLOW; Check level in affected SIG less than 16% lAW FRP-H.5, RESPONSE TO STEAM GENERATOR LOW LEVEL. The correct answer is B. A: Incorrect -No RED status, but correct actions for FRP-H.1. B: Correct -Any S/G pressure> 1140 is YELLOW status, entry condition for FRP-H.2, initial action is to ISOLATE flow to affected S/G. C: Incorrect -FRP-H.3 YELLOW would apply IF affected SIG Level was high. 0: Incorrect -Incorrect entry, SIG levels are required to be less than 8% for FRP-H.5 entry. Exam Question Number: 26

Reference:

FRP-H.2, Pages 3-4; H.1, Page 3; H.3, Page 3; H.5, Page 3; CSFST, Page 5. KA Statement: Knowledge of the interrelations between the (Steam Generator Overpressure) and the following: Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility. History: New -Written for HLC-08 NRC Exam. Tuesday, June 17, 20081 :21 :07 PM 30 Rev. FRP-H.2 RESPONSE TO STEAM GENERATOR OVERPRESSURE Page Purpose and Entry Conditions (Page 1 of 1) 1. PURPOSE This procedure provides actions for an overpressure condition affecting any Steam Generator where pressure has increased above the highest Steamline Safety Valve setpoint.

2. ENTRY CONDITIONS 8 3 CSF-3, Heat Sink Critical Safety Function Status Tree on a YELLOW condition.

-END -of 6 Rev. 8 FRP-H.2 RESPONSE TO STEAM GENERATOR OVERPRESSURE Page 4 of 6 INSTRUCTIONS RESPONSE NOT OBTAINED Throughout this procedure, "affected" refers to any S/G in which pressure is greater than 1140 psig. 1. Check S/G Pressures -ANY GREATER THAN 1140 PSIG 2. Verify FW Isolated To Affected S/G(s) :

• FW REG Valve(s) -CLOSED
• FW REG BYP Valve(s) -CLOSED
• FW HDR SECTION Valve(s) -CLOSED
• 3. Check Affected S/G(s) Level -LESS THAN 84% [82%] 4. Check APP-002-F7, INSTR AIR HDR LO PRESS -EXTINGUISHED

5. Dump Steam From The Affected S/G(s) Using One Of The Following Methods:

• STEAM LINE PORV(s)
• Unlock and close the breaker at MCC-8 for the affected S/G(s) and use the MSIV BYP(s) QR
• STEAM SHUTOFF(s) to SDAFW PUMP Reset SPDS bNQ return to procedure and step in efffect. Go To FRP-H.3, Response To Steam Generator High Level. Align nitrogen to Steam Line PORVs using AOP-017, Loss Of Instrument Air. Observe CAUTION prior to Step 7 bNQ Go To Step 7.

Rev. 22 FRP-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK Page 3 of Purpose and Entry Conditions (Page 1 of 1) 1. PURPOSE This procedure provides actions to respond to a loss of secondary heat sink in all Steam Generators.

2. ENTRY CONDITIONS

a. PATH-I, when m1nlmum AFW flow is not verified AND narrow range level in all S/Gs is less than 8% [18%]. b. CSF-3, Heat Sink Critical Safety Function Status Tree on a RED condition.

-END -45 Rev. FRP-H.3 RESPONSE TO STEAM GENERATOR HIGH LEVEL Page Purpose and Entry Conditions (Page 1 of 1) 1. PURPOSE This procedure provides actions to respond to a steam generator high level condition and to address the steam generator overfill concern. 2. ENTRY CONDITIONS

a. CSF-3, Heat Sink Critical Safety Function Status Tree on a YELLOW condition.

b. FRP-H.2, Response To Steam Generator Overpressure, if the affected SIG level is high. c. FRP-H.4, Response To Loss Of Normal Steam Release Capability, if the affected SIG level is high. -END -9 3 of 7 Rev. FRP-H.5 RESPONSE TO STEAM GENERATOR LOW LEVEL Page Purpose and Entry Conditions (Page 1 of 1) 1. PURPOSE This procedure provides actions to respond to a steam generator low level condition.

2. ENTRY CONDITIONS 7 3 CSF-3. Heat Sink Critical Safety Function Status Tree on a YELLOW condition.

-END -of 5 1 CSFST ENTER GREEN CSF-SAT NO NO NO CSF-3, HEAT SINK YELLOW GO TO FRP-H.5 YELLOW GO TO FRP-H.4 Rev. 4 NO YES YELLOW GO TO FRP-H.3 YELLOW GO TO FRP-H.2 RED GO TO FRP-H.1 Page 5 of 91 HLC-08 NRC Written Exam 27. Which ONE (1) of the following identifies the major concern or license limitation associated with post-LOCA Containment Flooding? A. ITS Bases for containment integrity are compromised in situations above maximum flood level. B. Dilution of sump water with CCW or Fire Water may potentially cause a return to criticality. C. Water introduced into the ECCS sump beyond capacity can potentially affect the operation of vital equipment. D. Loss of inventory from interfacing cooling systems can potentially result in a loss of ability to remove heat from the reactor core. 27 HLC-08 NRC Written Exam 27. WEIS G2.2.38 OOllCTMT FLOODING/1I2/3.6/4.5IROILOWININRNP AUDIT -2001/FRP-J.2-003 Which ONE (1) of the following identifies the major concern or license limitation associated with post-LOCA Containment Flooding? A. ITS Bases for containment integrity are compromised in situations above maximum flood level. B. Dilution of sump water with CCW or Fire Water may potentially cause a return to criticality. Water introduced into the ECCS sump beyond capacity can potentially affect the operation of vital equipment. D. Loss of inventory from interfacing cooling systems can potentially result in a loss of ability to remove heat from the reactor core. The correct answer is C. A: Incorrect -ITS Bases addresses Containment Integrity as part of OPERABILITY, but does NOT address Containment Flooding. B: Incorrect -Basis for FRP-J.2, CONTAINMENT FLOODING does NOT address reactivity as a concern. Both CCW and Firewater are isolated on Phase A and Phase B isolation. C: Correct -The purpose of the sump is to collect and divert water in areas that will NOT affect vital plant equipment. Flooding may jeopardize that function. D: Incorrect -A loss of ability to remove heat is dealt with in other procedures. The major concern in FRP-J.2 is containment flooding. Exam Question Number: 27

Reference:

FRP-J.2 BD, Page 3. KA Statement: Knowledge of conditions and limitations in the facility license. History: Tuesday, June 17, 2008 1 :21 :07 PM 31 DISCUSSION (From the WOG FR-Z.2 Basis Document)

1. INTRODUCTION Guideline FR-Z.2, RESPONSE TO CONTAINMENT FLOODING, is a Function Restoration Guideline (FRG) that provides procedural guidance when the containment level is greater than flood level. There is only one explicit transition to guideline FR-Z.2. It is from the Critical Safety Function Status Tree F-O.5, CONTAINMENT, on an ORANGE priority when containment sump level is greater than flood level. After all the actions in guideline FR-Z.2 are completed, the operator is instructed to return to the guideline and step in effect. 2. DESCRIPTION Guideline FR-Z.2, RESPONSE TO CONTAINMENT FLOODING, provides actions to respond when the containment level is greater than design flood level. This level is significant since the critical systems and components, which are necessary to ensure an orderly safe plant shutdown and provide feedback to the operator regarding plant conditions, are normally located above the design flood level. Therefore, the guideline FR-Z.2 is entered from the Containment Status Tree on an ORANGE priority when this design flood level is exceeded.

The primary purpose of the containment sump area is to collect the water injected into the containment or spilled from the reactor coolant system following an accident. The water collected in the containment sump is then available for long term core and/or containment cooling via the emergency core COOling or containment spray recirculation systems. In addition, the containment sump collects the injected or spilled water into areas such that vital systems or components will not be flooded and thus rendered inoperable. The maximum level of water in the containment following a major accident generally is based upon the entire water contents of the reactor coolant system, refueling water storage tank, condensate storage tank, and SI accumulators. This water volume approximates the maximum water volume introduced into the containment following a LOCA plus a steamline or feedline break inside containment. An indicated water level in the containment greater than the maximum expected volume (design basis flood level) is an indication that water volumes other than those represented by the above noted volumes have been introduced into the containment. Also, the high water level provides an indication that potential flooding of critical systems and components needed for plant recovery may occur. The actions in this guideline attempt to identify any unexpected source of water and isolate it if possible. Beyond that the plant engineering staff is consulted to determine if transfer of containment sump water to other tanks is appropriate.

3. RECOVERY/RESTORATION TECHNIQUE The objective of the recovery/restoration technique incorporated into guideline FR-Z.2 is to provide actions to respond to containment flooding.

The following subsections provide a summary of the major action categories of operator actions and the key utility decision points for guideline FR-Z.2, RESPONSE TO CONTAINMENT FLOODING. 3.1 High Level Action Summary A high level summary of the actions performed in FR-Z.2 is given below in the form of major action categories. These are discussed below in more detail. MAJOR ACTION CATEGORIES IN FR-Z.2 o Try to Identify Unexpected Source of Sump Water and Isolate It if Possible o Notify Plant Engineering Staff of Sump Level and Activity Level o Try to Identify Unexpected Source of Sump Water and Isolate It if Possible The first action in this guideline is to try to identify the source of water which is causing containment flooding and isolate it. The concern regarding flooding is that critical plant components needed for plant recovery could be damaged and rendered inoperable. o Notify Plant Engineering Staff of Sump Level and Activity Level By knowing the sump level and activity level, the plant engineering staff can determine if the excess water can be transferred to storage tanks located outside containment. I FRP-J.2-BD Rev 3 Page 3 of 6/

1. E15 EK1.2 001ll11111!

QUESTIONS REPORT for AUDIT Given the following conditions:

• A LOCA has occu rred.
• Due to an abnormal rise in containment sump level, the crew has transitioned to FRP-J.2, Response to Containment Flooding.

Which ONE (1) of the following identifies the major concern associated with the actions in this procedure? A. Dilution of sump water may potentially cause a post-LOCA reactivity transient resulting in loss of subcriticaJity. B. Contaminants in water from other systems can potentially block flow channels during the long term cooling phase of the recovery. Water introduced into the sump beyond the capacity to contain it in appropriate areas can potentially affect the operation of vital equipment. D. Loss of inventory from interfacing cooling systems can potentially result in a loss of ability to remove heat from the reactor Core. A. Incorrect. Procedure does not consider reactivity abnormally. B. Incorrect. Other systems may contain contaiminants but post-LOCA not a major threat to cooling. C. Correct. The purpose of the sump is to collect and divert water in areas that will not affect vital plant equipment. Flooding may jeopardize that function. D. Incorrect. The major concern is flooding for this procedure. Loss of ability to remove heat is dealt with in other procedures. Common Question 064 Tier 1 Group 2 KIA Importance Rating -RO 23.7/ SRO 2.7 Knowledge of the operational implications of the following concepts as they apply to the (Containment Flooding) Normal, abnormal and emergency operating procedures associated with (Containment Flooding). 10 CFR Part 55 Content -55.41 Comments -Category 1: Category 3: Category 5: Category 7: Thursday, June 05, 2008 1 :44:01 PM Category 2: Category 4: Category 6: Category 8: 1 HLC-08 NRC Written Exam 28. During containment isolation valve testing, CCW-716A, CCW to RCP ISOLATION VALVE, was inadvertently CLOSED with RCP "B" running. Which ONE (1) of the following RCP components suffered a loss of cooling water flow? A. Motor Bearing Oil Coolers ONLY. B. Thermal Barrier Heat Exchangers ONLY. C. Thermal Barrier Heat Exchangers and Motor Bearing Oil Coolers. D. Motor Air Coolers, Thermal Barrier Heat Exchangers, and Motor Bearing Oil Coolers. 28 HLC-08 NRC Written Exam 28. 003 K6.04 OOllREAC COOL PUMP/2/1/2.S/3.1IROILOW/N/A/NEW -200S/CCW-009 During containment isolation valve testing, CCW-716A, CCW to RCP ISOLATION VALVE, was inadvertently CLOSED with RCP "B" running. Which ONE (1) of the following RCP components suffered a loss of cooling water flow? A. Motor Bearing Oil Coolers ONLY. B. Thermal Barrier Heat Exchangers ONLY. Ct Thermal Barrier Heat Exchangers and Motor Bearing Oil Coolers. D. Motor Air Coolers, Thermal Barrier Heat Exchangers, and Motor Bearing Oil Coolers. The correct answer is C. A: Incorrect -Motor Oil Cooler is correct, but NOT a complete answer as Thermal Barrier HX is also cooled by CCW. B: Incorrect -Thermal Barrier HX is correct, but NOT a complete answer as the Motor Oil Cooler is also cooled by CCW. C: Correct -CCW -716A is the CCW supply to all RCP cooling. Discharge lines have seperate isolation valves. 0: Incorrect -Motor Air Coolers are CV HVH units which are cooled with Service Water. Exam Question Number: 28

Reference:

SD-013, CCW, Pages 20-21, Figure 3. KA Statement: Knowledge of the effect of a loss or malfunction on the following will have on the RCPS: Containment isolation valves affecting RCP operation. History: New -Written for HLC-08 NRC exam. Tuesday, June 17,2008 1 :21 :07 PM 32 SD-013 COMPONENT COOLING WATER SYSTEM LCV-I030A supplies auto makeup from the CCW system to WGC "A". Level is automatically controlled by makeup and drain valves in conjunction with a level controller on each separator. Local push buttons (seal water bypass switch [es]) allow for manual filling prior to WGC start (Push buttons bypass running interlocks). A low level (7.5") will open the separator's makeup valve from CCW. 5.4 LCV-1032A, Waste Gas Compressor "B" Seal Water Level Control LCV-1032A supplies auto makeup from the CCW system to the WGC "B". Level is automatically controlled by makeup and drain valves in conjunction with a level controller on each separator. Local push buttons (seal water bypass switch(es>> allow for manual filling prior to WGC start (Pushbuttons bypass running interlocks). A low level (7.5") will open the separator's makeup valve from CCW. 5.5 TCV-144, Non-Regenerative Heat Exchanger Temperature Control Valve (CWD-B-190628 Sh00471) The flow of CCW through the non-regenerative heat exchanger is controlled by TCV-144, which is controlled by the Temperature Controller TC-144 located on the RTGB. The temperature signal is provided by TE-144 located in the CVCS downstream of the non-regenerative heat exchanger. TC-144 is a 0-10 turn potentiometer. The controller has both Automatic and Manual control capability. The 0-10 turn potentiometer is used to adjust the control set point for TCV -144 when operated in Automatic. Pushbuttons located on the controller are used to transfer from Automatic control to Manual control and vice-versa. When in Manual control the operator opens or closes the valve by depressing OPEN or CLOSE push buttons located on the RTGB mounted controller. Valve TCV-144 is located on the 2nd level ofthe Non-Regenerative Heat Exchanger Room. , 5.6 ) CC-716A1B, CCW to RCP Isolation -Motor Operated Valve(MOV) \..---"" ccw (CWD-B-190628 Sh00231, Sh00232) Valves CC-716A and CC-716B are operated by two position (OPEN/CLOSE), spring return to center, switches located on the RTGB. The valves also automatically close on a "P" signal (CS) providing containment isolation of this potential leak path. The valves are located outside containment downstream of CV Penetration P-18 in the Auxiliary Building pipe alley. The valves are powered from MCC-5 and 6 respectively. Valve position indication is provided on the RTGB at the control switch and as part of the Page 20 of38 Revision 9 INFORMATION USE ONLY SD-013 COMPONENT COOLING WATER SYSTEM Containment Phase B isolation indications. 5.7 CC-730, Bearing Outlet Isolation -MOV (CWD-B-190628 Sh00233) Valve CC-730 is operated by a two position (OPEN/CLOSE), spring return to center, switch located on the RTGB. The valve also automatically closes on a UP" signal providing containment isolation of this potential leak path. The valve is located outside containment downstream ofCV Penetration P-19 in the Auxiliary Building pipe alley. The valve is powered from MCC-6. Valve position indication is provided on the RTGB at the control switch and as part of the Containment Phase B isolation indications. 5.8 FCV-626, Thermal Barrier Outlet Isolation Flow Control-MOV (CWD-B-190628 Sh00234) Valve FCV-626 is operated by a two position (OPEN/CLOSE), spring return to center, switch located on the RTGB. The valve will automatically close on a high flow (100 gpm) as monitored by FIC-626. The valve also automatically closes on a "P" signal providing containment isolation of this potential leak path. The valve is located outside containment downstream of CV Penetration P-20 in the Auxiliary Building pipe alley. The valve is powered from MCC-6. Valve position indication is provided on the RTGB at the control switch and as part of the Containment Phase B isolation indications. 5.9 CC-735, Thermal Barrier Outlet Isolation -MOV (CWD-B-190628 Sh00230) Valve CC-735 is operated from the two position (OPEN/CLOSED), spring return to the center, switch located on the RTGB. The valve is located outside containment downstream of FCV-626 in the Auxiliary Building pipe alley. The motor operator for the valve is powered from MCC-5. Upon receipt ofa "P" signal, the valve will close. 5.10 CC-739, Excess Letdown HX. Outlet Isolation Valve, Air Operated Valve ccw (CWD-B-190628 Sh00229) Valve CC-739 is operated from the RTGB using a two position (OPEN/CLOSED), spring return to center, switch. This valve is located in the Auxiliary Building pipe alley and provides CV isolation downstream of CV penetration P-22. CC-739 is an Air Operated Valve that receives operating air from the instrument air system through 125V DC solenoid valves. The solenoid valves receive power from the 125V DC Auxiliary Panel GC CKT#29. A safeguards actuation signal, "T" signal, will de-energize the Page 21 of38 Revision 9 INFORMATION USE ONLY COMPONENT COOLING WATER SYSTEM CCW-FIGURE-3 SPENT FUEL PITHX WASTE EVAPORATOR "A" r----l CONTROL ROD DRIVE COOLERS RHR PUMPS r EXCESS LETDOWN HEAT EXCHANGER ..... a-f"'...,.,..--------I REACTOR __ PUMPS L INSIDE CONTAINMENTJ

---

OUTSIDE CONTAINMENT HLC-08 NRC Written Exam 29. Given the following: -GP-002, COLD SHUTDOWN TO HOT SUBCRITICAL AT NO LOAD T AVG' is in progress. -RHR has been removed from service. -RCS heatup is in progress. -RCS is at 370 OF and 800 PSIG. -The RO notes that RCS pressure is no longer increasing. -VCT Makeup is in progress. Which ONE (1) of the following plant conditions has caused the RCS pressure response, and what is (are) the parameter(s) the crew can use to verify the cause? A. S/G tube leakage; monitor increasing level in the S/G with feed flow secured. B. RHR-706, RHR SYSTEM RELIEF VALVE lifting; monitor PRT level and pressure. C. CVC-209, LOW PRESSURE RELIEF VALVE lifting; monitor letdown line flow and temperature. D. HCV-121 has closed, isolating charging line; monitor FI-122A, CHARGING LINE FLOW INDICATOR. 29 HLC-08 NRC Written Exam 29.004 A1.0S 001lCVCS/2/112.9/3.2IROIHIGHlNIAlNEW -2008/CVCS-008 Given the following: -GP-002, COLD SHUTDOWN TO HOT SUBCRITICAL AT NO LOAD T AVG' is in progress. -RHR has been removed from service. -RCS heatup is in progress. -RCS is at 370 of and SOO PSIG. -The RO notes that RCS pressure is no longer increasing. -VCT Makeup is in progress. Which ONE (1) of the following plant conditions has caused the RCS pressure response, and what is (are) the parameter(s) the crew can use to verify the cause? A'I S/G tube leakage; monitor increasing level in the S/G with feed flow secured. B. RHR-706, RHR SYSTEM RELIEF VALVE lifting; monitor PRT level and pressure. C. CVC-209, LOW PRESSURE RELIEF VALVE lifting; monitor letdown line flow and temperature. D. HCV-121 has closed, isolating charging line; monitor FI-122A, CHARGING LINE FLOW INDICATOR. The correct answer is A. A: Correct -At 370 of, S/G pressure is -160 PSIG. Any S/G tube leakage will result in RCS to S/G leakage. B: Incorrect -RHR system has been isolated in the question stem. C: Incorrect -CVC-2091ifts to the VCT. Valve lifting will divert flow from the demineralizers but will NOT stop flow. 0: Incorrect -RCP Seal Injection flow would continue via seal injection lines. Exam Question Number: 29

Reference:

GP-002, Section S.4.7, Page 65,66; SD-021 CVCS, Pages 37,38, Figure 1, SD-003 RHR, Figure 6. KA Statement: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CVCS controls including: S/G pressure and level. History: New -Written for HLC-OS NRC exam. Tuesday, June 17, 2008 1 :21 :08 PM 33 S.4.7 (Continued) CAUTION If the starting limitations stated in the Precautions and Limitations Section of OP-201 are exceeded, motor damage can occur due to motor overheating. NOTE: When an RHR pump is started and its train ventilation unit is inoperable, the opposite train ventilation can be started by placing the RUN/AUTO switch on the power supply breaker to RUN. I I GP-002 3. Alternately operate RHR Pump "A" and "8" to -maintain the outlet temperatures, as read on TR-604 Pens 1 and 3, within 50°F until the temperatures are less than 125 ° F. (CR 95-00565) -TR-604 Pen 1 temperature less than 125°F of -TR-604 Pen 3 temperature less than 125°F of 4. Stop the RHR Pumps AND verify both RHR pump room ventilation units are STOPPED. -RHR Pump "A" -RHR Pump "8" -HVH-SA -HVH-S8 Perform the following:

a. Close RHR-750, LOOP 2 HOT LEG TO RHR SYSTEM. b. Close RHR-751, LOOP 2 HOT LEG TO RHR SYSTEM. c. d. Open the breaker for RHR-750, LOOP 2 HOT LEG TO RHR SYSTEM, on MCC-5 in CMPT NO. 12C. Open the breaker for RHR-751, LOOP 2 HOT LEG TO RHR SYSTEM, on MCC-6 in CMPT NO. SM. Rev. 103 Page 65 of 1241 8.4.7 (Continued)

6. Adjust PC-145, PRESSURE, to increase Letdown pressure to within 25 psig of current RCS Pressure.

7. Adjust HIC-142, PURIFICATION FLOW, to open HCV-142, RHR TO LETDOWN LINE. 8. Adjust PC-145 to decrease letdown pressure to greater than 140 psig and less than 210 psig. 9. As Letdown increases, adjust PC-145 setting OR isolate letdown orifices to maintain Letdown flow below 120 gpm. 10. Control Charging pump speed, letdown flow and excess letdown flow to maintain PZR level between 30% and 40%. NOTE: Leaving HCV-142, RHR TO LETDOWN LINE, open until the RHR System is less than 210 psig will allow SI-862A and SI-8628, RWST TO RHR valves, to open. I GP-002 11. WHEN RHR System pressure is less than 210 psig as indicated on PI-602A and PI-6028, THEN adjust HIC-142, PURIFICATION FLOW, to close HCV-142. 12. Close RHR-760, RHR SYSTEM TO LETDOWN LINE. (ACR 94-00533)

13. Adjust HIC-758, RHR HX DISCH FLOW to 0% demand. 14. Open the RWST to RHR Pump Suction Valves AND record time. -SI-862A, RWST TO RHR -SI-8628, RWST TO RHR Time __ _ Rev. 103 Page 66 of 1241 SD-021 CHEMICAL AND VOLUME CONTROL SYSTEM Pressure Letdown Valve (PCV -145) and Low Pressure Letdown Relief Valve L/<CVC-209)

\ This valve is controlled from the RTGB in either automatic or manual to keep the water in the piping downstream of the orifices to PCV -145 from flashing to steam. During normal plant operation this valve is adjusted to maintain approximately 300 psig upstream of the valve. For plant heatups and cooldowns the valve is adjusted to maintain the proper system pressure for evolutions in progress; i.e., running RCPs, etc. PCV-145 is air operated and fails open. The pressure signal used to operate this valve in automatic comes from PT -145, located upstream of the valve. PT-145 drives pressure indicator PI-145 located on the RTGB. The controller for PCV-145, located on the RTGB, consists of a dial potentiometer for automatic setpoint control, pushbuttons for manual control and a controller output demand indication. Opening the valve results in a lower backpressure on PT -145 and closing the valve results in a higher backpressure. PT-145 alarms APP-OOI-D6, LP LTDN LN HI.PRESS, at 400 psig. Low Pressure Letdown Relief Valve, CVC-209, is located just downstream of PC V-145. CVC-209 has a setpoint of 200 psig and relieves to the VCT. 5.1.5 VCT/Demineralizer Diversion Valve (TCV-143) TCV-143 is a three way valve that allows the operator to bypass the demineralizers. It is controlled by a three position (VCT, AUTO, DEMIN) RTGB switch which is normally selected to AUTO. The demineralizers will be automatically bypassed if letdown temperature reaches a pre-determined setpoint (135°F). The demineralizers will be automatically un-bypassed when temperature falls to 130 degrees. The temperature used for this function is sensed by TE-143. TE-143 also provides the input for a temperature indicator on the RTGB. The valve fails to the VeT position. 5.1.6 VCT/Deborating Demineralizer Diversion Valve (CVC-244) This solenoid-operated valve allows the operator to place the deborating demineralizer in service when needed. It is controlled by a two position (VCT/DEB DEMIN) RTGB switch. It is selected to VeT except at end of cycle. CVC-244 fails to the VeT position. More information on CVC-244 operation can be found in section 3.4.3 of this System Description. 5.1.7 VCT Vent Valves (CVC-258 and PCV-117) eves The VCT gas space can be vented to the Waste Disposal Gas System by operating the solenoid valve CVC-258. CVC-258 is controlled by a OPEN/CLOSE RTGB switch. PCV-117 senses VCT pressure and will shut if VCT pressure is reduced to 15 psig Page 37 of71 Revision 1 0 INFORMATION USE ONLY SO-021 CHEMICAL AND VOLUME CONTROL SYSTEM which is required for proper RCP seal operation. 5.1.8 Hydrogen and Nitrogen Pressure Regulating Valves (PCV-118 and PCV-119) Hydrogen gas is supplied to the VCT during normal operations. The H2 pressure is regulated to approximately 22-28 psig by PCV-118. During shutdown operations nitrogen gas is supplied to the VCT through PCV-119 which regulates the N2 pressure to approximately 22-28 psig. 6 Charging Flow Yalve (HCY -121) , This air operated valve is normally open and is controlled by a RTGB potentiometer. \\) Throttling with this valve will increase backpressure on the charging line and cause the -4 RCP seal flow to increase. HCV-12l fails open. ' CAUTION HCV-12l is normally full open. Extra care should be taken when throttling due to the charging pumps being positive displacement pumps; rapid pressure increase will occur if a proper flow path is not maintained. 5.1.10 Seal Water Return Isolation Valve (CVC-381) A motor operated valve is provided. It is normally open and will shut on a Phase "B" containment isolation signal. 5 .1.11 Charging Line Isolations Three return paths to the RCS are available on the charging line. Since no relief valve is provided on this section of the charging line, a locked open manual valve (CVC-312) bypasses CVC-31OA. This valve should not be closed except by specific procedural control. 5.1.11.1 Charging Line to Loop 1 Hot Leg (CVC-31OA) Air operated valve, CVC-31OA, is normally closed but may be open to provide an alternate flow path to the RCS from the charging line. It is controlled from the RTGB by a two position (OPEN/CLOSE) switch. CVC-31OA fails open. 5.1.11.2 Charging Line to Loop 2 Cold Leg (CVC-310B) Air operated valve, CVC-31OB, is normally open to provide a flow path to the RCS. It eves Page 38 of71 Revision 1 0 INFORMATION USE ONLY TOPRT* .. -r-....., LOOPl LOOPl HOTLEG..., ............... AUX SPRA Y4lA Il'. I LOOP 2 COLD LEG From Loop 2 Cold Leg REGEN \ HX EXCESS LTDN HX 389 From RCP Seals eves FLOW DIAGRAM CVCS-FIGURE-l 320 Charging Pulsation Pumps Dampener 286 " To Rep Seals JI-<xl-"'---< INFORMATION USE ONLY From RWST 227 Deborating Demins. 237A tel 352 353 223 Mixed Bed Demins. 225 216 Cation Bed Demin. FromBA Transfer Pumps TOSI PUMP 8&C * .. SUCTIONS I SI*891 C RHR*764 SI-8910 RHR*7448 RHR SYSTEM -NORMAL (AT POWER) LINEUP RHR-FIGURE 6 '!" RHR-7548 SI-887 RHR PUMP 8 TO SIANO CONTAINMENT SPRAY PUMP SUCTIONS .... o SI*8768 SI*8758 '!" '1-+--f><1-l--I RHR*754A RHR*757A TO RC LOOP 1 COLD LEG TO RC LOOP 3 COLD LEG TO RCLOOP2 COLD LEG MINIFLOW RECIRC RHR-743 RHRPUMPA -. FROM CONTAINMENT SUMP RHR HEAT-UP LINE INFORMATION USE ONLY !l! :::" !l! :::" .... U1 o FROM RC LOOP 2 HOT LEG HLC-08 NRC Written Exam 30. Given the following: -A Small Break LOCA occurred 3 hours ago. -RCS pressure is 210 PSIG. -The alignment of the ECCS for Cold Leg Recirculation is complete lAW EPP-9, TRANSFER TO COLD LEG RECIRCULATION. Which ONE (1) of the following describes the current RHR/ECCS alignment? ONE (1) RHR Pump and ONE (1) SI Pump operating in series (Piggy-back mode), with the ... A. RHR AND SI Pump suctions from the CV sump, discharging to RCS cold legs via RHR HX (RHR Pump), and the BIT. B. RHR Pump suction from the CV sump, discharging through RHR HX to SI Pump suction, SI Pump discharging to RCS cold legs via the BIT. C. RHR Pump suction from the CV sump, discharging through RHR HX to RCS cold legs AND to SI Pump suction, SI Pump discharging to RCS cold legs via the BIT. D. RHR Pump suction from the CV sump, discharging through RHR HX to SI Pump suction, SI Pump discharging to RCS hot legs. 30 HLC-08 NRC Written Exam 30. 005 Kl.06 001/RHR/2/1/3.5/3.6/RO/HIGHJN/A/NEW -2008/RHR-008 Given the following: -A Small Break LOCA occurred 3 hours ago. -RCS pressure is 210 PSIG. -The alignment of the ECCS for Cold Leg Recirculation is complete lAW EPP-9, TRANSFER TO COLD LEG RECIRCULATION. Which ONE (1) of the following describes the current RHR/ECCS alignment? ONE (1) RHR Pump and ONE (1) SI Pump operating in series (Piggy-back mode), with the ... A. RHR AND SI Pump suctions from the CV sump, discharging to RCS cold legs via RHR HX (RHR Pump), and the BIT. B:I RHR Pump suction from the CV sump, discharging through RHR HX to SI Pump suction, SI Pump discharging to RCS cold legs via the BIT. C. RHR Pump suction from the CV sump,4discharging through RHR HX to RCS cold legs AND to SI Pump suction, SI Pump discharging to RCS cold legs via the BIT. D. RHR Pump suction from the CV sump, discharging through RHR HX to SI Pump suction, SI Pump discharging to RCS hot legs. The correct answer is B. A: Incorrect -No CV sump suction to the SI pumps. B: Correct -This is the proper alignment and injection path due to RCS pressure being greater than the RHR Pump shutoff head (110 PSIG). C: Incorrect -RCS pressure is above RHR shutoff head. 0: Incorrect -Correct alignment for Hot leg recirculation 11 hours after the accident. Exam Question Number: 30

Reference:

SD-002, SI, Pages 9, 10, 11, Figure 5. KA Statement: Knowledge of the physical connections and/or cause-effect relationships between the RHRS and the following systems: ECCS. History: New -Written for HLC-08 NRC exam. Tuesday, June 17,20081:21:08 PM 34 SD-002 SAFETY INJECTION SYSTEM 2.3 For any rupture of a steam pipe and the associated uncontrolled heat removal from the core, the SI system adds shutdown reactivity so that, with a stuck rod, no offsite power, and minimum engineered safety features, there is no consequential damage to the RCS and the core remains in place and intact. Redundancy and segregation of instrumentation and components are incorporated in the design to assure that postulated malfunctions will not impair the ability of the system to meet the design objectives. The system is effective in the event of loss of normal plant auxiliary power coincident with the loss of coolant, and can accommodate the credible failure of any single component or instrument channel to respond actively in the system, During the recirculation phase of a LOCA, the system can accommodate a loss of any part of the flow path, since backup alternate flow path capability is provided. Pipe whip protection for the ECCS components is provided as is protection against seismic events, protection against missiles, and loads that may result from the effects of a LOCA. The accumulators, which are passive components, discharge into the cold legs of the reactor coolant piping when RCS pressure decreases to 630-640 psig, thus assuring rapid core cooling for large pipe breaks. They are located inside the containment, but outside the crane wall. Therefore each accumulator is protected against possible missiles. System Flow .. 2.3.1 Injection Phase SI When in the Injection Phase, there are three modes of component operation. The modes are High Head/Low Flow Active Injection (SI Pumps), Passive Accumulator Injection and Low Head/High Flow Active Injection (RHR Pumps). NOTE: One, two or all three modes may be in operation at any given time. (SI-Figure-1) Upon the initiation of High Head/Low Flow Active Injection Mode, the SI Pumps take a suction on the RWST through SI-864A and SI-864B and inject into the cold legs via pump discharge cross-connect valve SI-878A and SI-878B, SI-867 A and SI-867B, the Boron Injection Tank, SI-870A and SI-870B. During Passive Accumulator Injection the pressurized nitrogen blanket will automatically discharge water from the Accumulator into the cold legs. Page 9 of 43 Revision 14 INFORMATION USE ONLY SD-002 SAFETY INJECTION SYSTEM During Low Head/High Flow Active Injection, the RHR Pumps take a suction from the RWST via SI-864A and SI-864B, SI-862A and SI-862B, RHR-752A and RHR-752B. From the RHR pumps the borated water flows through the RHR heat exchangers, RHR-744A and RHR-744B and into the RCS cold legs. This path can be traced on Figure-2. 2.3.2 Recirculation Phase During any mode of the injection phase, if the RWST volume of borated water decreases to 27 % an operator will switch from the Injection Phase to the Recirculation Phase. The source of borated water will be changed from the RWST to the CV Sump. The SI and CV spray pumps can not take a suction from the CV sump. In order to obtain CV spray or SI Pump operation when RWST inventory is depleted, the pumps must be aligned in what is commonly referred to as the "Piggy Back" Mode. In this mode of operation an RHR pump takes a suction on the containment floor and discharges to the suction of the SI and CV Spray Pumps. 2.3.2.1 RHR Flow> 1200 gpm (SI-Figure-2) RHR pumps take a suction from the CV sump via SI-860A and SI-861A for RHR Pump "A" and SI-860B and SI-861B for RHR Pump "B". From the RHR pumps the borated water flows through the RHR heat exchangers, RHR-744A and RHR-744B and into the RCS cold legs.(Reference EPP-9) 2.3.2.2 RHR Flow < 1200 gpm (SI-Figure-3) RHR pumps take suction from the CV sump as described above. From the RHR pumps the CV sump water is supplied to the SI Pumps suction via SI-863A or SI-863B. The path down stream of the SI pumps is the same as for the Injection Phase. (Reference EPP-9) 2.3.3 Long Term Recirculation Phase with RCS pressure < 125 psig (SI-Figure-4) SI RHR pumps take suction from the CV sump. From the RHR pumps the CV sump water is supplied to the SI Pumps suction via SI-863A or SI-863B. The SI pumps will discharge to the RCS hot legs (loop 2 & 3) via SI-869, and then through SI-866A or SI-866B. The RHR pump discharge will be throttled through RHR-759 A & B to the RCS cold legs via RHR-744A and RHR-744B. (Reference EPP-lO) ong Term Recirculation Phase with ReS pressure> 125 psig (SI-Figure-5) Page 10 of 43 Revision 14 INFORMATION USE ONLY SD-002 SAFETY INJECTION SYSTEM \ 2.4 SI RHR pumps take suction from the CV sump. From the RHR pumps the CV sump water is supplied to the SI Pumps suction via SI-863A or SI-863B. The path for cold leg injection down stream of the SI pumps is about the same as for the Injection Phase into the cold legs. When in hot leg recirculation the SI pump discharge is aligned to supply water to the RCS hot legs loop 2 and loop 3 via SI-869, and then through SI-866A or SI -866B. Hot leg and cold leg injection from the SI pumps is alternated every 16 hours. System Description The SI System is comprised of three sources of borated water (RWST, Accumulator, and CV Sump) and the associated flow path necessary to deliver borated water to the Reactor Core. Some major components of the SI System and their locations are:

• RWST -north of the Auxiliary Building
• SI Pumps-northwest Auxiliary Building, elevation 226'
• Accumulators

-CV, one north, one north-northwest, and one south-southwest elevation 251.5' Under controlled plant conditions and if the Reactor Coolant volume is at an acceptable level, the SI System is in a stand-by or a shutdown condition. However, under a LOCA condition, the SI System will automatically align appropriate flow paths to insure flow from the borated water sources to the Reactor core. A signal which initiates operations for the Injection Phase is generated in one of six ways: (Refer to SD-006, Safeguard System Description).

1. 2. 3. 4. 5. 6.

• 2-out-of-3 Low Pressurizer Pressure (1715 psig)* 2-out-of-3 High Steam Line Differential Pressure (100 psid, PH -PL on 1/3 lines)
• High Steam Flow from 2-out-of-3 S/G's (37.25% to 109%) coincident with Low Tavg in 2-out-of-3 loops (543 OF) or Low Steam Pressure in 2-out-of-3 S/G's (614 psig) ** 2-out-of-3 High Containment Pressure (4 psig) 2-out-of-3 on 2-out-of-2 Hi-Hi Containment Pressure (10 psig) (Initiates Spray and SI) ** Manual actuation Automatic initiation may be manually blocked when the plant is below 2000 psig Page 11 of43 Revision 14 INFORMATION USE ONLY LONG TERM RECIRCULATION, RCS PRESSURE>

125 PSIG OR ONLY 1 RHR PUMP A V AILABLE SI-FIGURE-5 51-863MB opeN SI ACCUMULATORS OP 1,2,3 COLD lEGS RHR-759A RHR-744A&B ALTERNATE BETWEEN HOT LEG AND COLD LEG INJECTION EVERY 16 HOURS INFORMATION USE ONLY RWST RHR-751 RHR-750 LOOP 2 HOTLEG HLC-08 NRC Written Exam 31. The plant is operating at 100% RTP with a normal electrical lineup. -Breaker 52/10, 4kV Bus 1-2 Tie Bkr, trips OPEN. -An inadvertent 8afety Injection signal occurs. Which ONE (1) of the following describes the source of power to the 81 Pumps? A. 81 Pumps "A" and "e" are powered from off-site power. B. 81 Pump "A" is powered from EDG "A", 81 Pump "e" is powered from off-site power. C. 81 Pump "A" is powered from off-site power, 81 Pump "e" is powered from EDG "B". D. 81 Pumps "A" and "e" are powered from EDG "A" and "B", respectively. 31 HLC-08 NRC Written Exam 31. 006 K2.0l OOllEMERG CORE COOLING/2/113.6/3.9IROIHIGB/N/NNEW -2008/S1-006 The plant is operating at 100% RTP with a normal electrical lineup. -Breaker 52110, 4kV Bus 1-2 Tie Bkr, trips OPEN. -An inadvertent Safety Injection signal occurs. Which ONE (1) of the following describes the source of power to the SI Pumps? A. 81 Pumps "A" and "c" are powered from off-site power. 81 Pump "A" is powered from EDG "A", SI Pump "c" is powered from off-site power. C. 81 Pump "A" is powered from off-site power, 81 Pump "c" is powered from EDG "B". D. 81 Pumps "A" and "c" are powered from EDG "A" and "B", respectively. The correct answer is B. A: Incorrect -When breaker 52/10 trips, 4kV Bus 2 de-energizes, this de-energizes 480V Bus 1 and E-1. EDG "A" starts and energizes 480V Bus E-1. ONLY E-1 is energized from the EDG B: Correct -When breaker 52/10 trips, 4kV Bus 2 de-energizes, this de-energizes 480V Bus 1 and E-1. EDG "A" starts and energizes 480V Bus E-1. Bus E-2 remains energized from off-site power. C: Incorrect -EDG "B" starts on SI signal but does NOT energize 480V Bus E-2. D: Incorrect -EDG "B" starts on SI signal but does NOT energize 480V Bus E-2. Exam Question Number: 31

Reference:

OP-603, Page 15; GP-005, Page 54, 8ection 8.4.42; SD-039 KVAC, Pages 12-14, Figure 2. KA Statement: Knowledge of bus power supplies to the following: ECCS pumps. History: New -Written for HLC-08 NRC exam. Tuesday, June 17, 2008 1 :21 :08 PM 35 REFERENCE USE Section 8.1.2 Page 1 of 2 8.1.2 Transfer of Bulk Auxiliary Electrical Load from Startup Transformer to Unit Auxiliary TC "Transfer of Bulk Auxiliary Electr'ical Load from Startup Transformer to Unit Auxiliary Transformer" \f C \1 "3" } 8.1.2.1. Initial Conditions NOTE: This section of OP-603 has been screened lAW PLP-037 criteria and determined to be a Case Three activity. No additional management involvement is required beyond that routinely provided by first line supervision. CAUTION All clearances associated with this operation shall be canceled and tags removed before placing the system into operation.

1. Unit 2 is synchronized with the system AND is between 90 and 110 MWe lAW GP-005. 2. The Unit Auxiliary AND Main Bank Transformers cooling fans AND oil pumps are running AND have been checked. 3. The Bus Duct Cooling Fans are running AND have been checked. 8.1.2.2. Instructions CAUTION Extreme caution shall be taken to ensure all breaker interlocks function as designed.

IOP-603 1. INSERT key into UNIT AUX TRANSF synchroscope switch AND TURN switch to UNIT AUX BUS 1 pOSition. I Rev. 78 Page 15 of 165/ NOTE: Power Ramp Rate Limits are restricted after core fuel movement to 3.5%/hr from 50% to 100% power. During subsequent power increases, this ramp limit may apply depending on the maximum power level achieved and length of operation at that power level. (ESR 98-00395) (SOER 90-2, Rec. 2C) CAUTION If Power Ramp Rate Restrictions are in effect, then Power shall be increased based on the highest indication of Reactor Power. 8.4.40 IF all indications of Reactor Power agree within 5% of each other AND Turbine Control is in OPER AUTO, OR management approval has been obtained, THEN CONTINUE the load increase as follows: 1. IF the power ramp rate restrictions are in effect, THEN RECORD the rate limitations AND power limits: ___ %/hr from __ % to __ % rated power 2. ADJUST the SETTER indication using the REF v and/or REF A pushbuttons to indicate NO greater than 30.0 load. 3. DEPRESS the GO and/or HOLD pushbuttons AND the REF V and/or REF A. as necessary to continue the load increase to 30% Reactor Power, OR as directed by the Reactor Engineer. 8.4.41 IF all indications of Reactor Power agree within 5% of each other AND Turbine Control is in TURB MANUAL, THEN INCREASE load using the GV A. to 30% Reactor Power, OR as directed by the Reactor Engineer. the Generator is carrying between 90 MWe and 110 MWe, . TRANSFER the Bulk Auxiliary Electrical Load from StartlJQ,. ---Transformer to Unit Auxiliary Transformer lAW OP-603 while continuing withthis procedure. I GP-005 Rev. 85 Page 54 of 70 I SD-039 230/4 KV AC ELECTRICAL SYSTEM The primary windings of the UAT are delta connected while the secondary is two capacity wye connected windings. Each secondary winding is grounded through an inductor. 3.6 SUT MFG TYPE RATING Primary Windings: Secondary Windings: Voltage Ratio & Connections Westinghouse FOA H=44MVA, 3 PHASE , 60HZ, FOA, 55"C Rise, 49.2MVA FOA, 65"C Rise 120.75/117.875/115.00/112.125/109.250 KV SET ON 115.000KV X,Y= 22MVA FOA 55°C Rise, 24. 640MV A 65"C Rise, 4.368KV 115 KV Wye -2 half capacity 4.368 KV Wye The SUT is a three phase step-down transformer. The primary side of this transformer is connected to the Unit 1 115KV Span Bus. The secondary voltage is reduced to 4368 volts (name plate rating). This transformer is used at all times and normally carries all plant loads when the unit is off-line. The windings are wye-wye connected. This transformer is also filled with oil. There are four circulating pumps and cooling coils associated with this transformer. There is one fan for each cooling coil. Oil filled bushings are also used to insulate the electrical cables from the casing. There is a set of motor operated disconnects on the supply side of this transformer to isolate it from the 115KV Span Bus. The disconnects are to only be opened when there is no load on the transformer. The motor is a DC motor powered from "A" Station Battery and is operated from the RTGB. A means of disconnecting the motor and manual operation is also provided. When the generator is not on-line, all five 4160V buses are powered from this transformer, unless we are using back feed through the U AT during maintenance on the SUT. With the generator on-line and the UAT in-service, only 4160V Bus 3 will be supplied from this transformer. (' 3.7 .)4160V Buses (see Figures 5 &6) ' ............ ! There are five (5) medium voltage (4KV) buses that supply power to the major plant components, 4KV motors and seven (7) SSTs that provide the plant's low voltage , service. These buses contain current transformers (CTs), PTs and other components that provide signals for local metering, RTGB metering, protection devices that detect KVAC Page 120f39 Revision 14 INFORMATION USE ONLY SD-039 230/4 KV AC ELECTRICAL SYSTEM over current, over voltage for grounds, under voltage (UV) and under frequency (UF). Annunciation in the Control Room is provided for 4KV motor over current conditions, 4KV UV conditions, 4KV UF conditions, and 4KV cooling fan failure. 4KV pump motor, SST and incoming supply breakers have current indication on their respective cubicles that receive inputs from CTs on the supply cables. The PT and Fan cubicles provide voltage indication. The loads supplied by each bus can be found in Electrical Distribution Procedure EDP-001. (See Figure 2) The 4160V system is supplied via the UAT andlor the SUT, each of which have two secondary windings. 4KV Buses 1 and 4 can be supplied from the UAT via 4KV breakers 5217 and 52120 respectively. 4KV Bus 5 is supplied from 4KV Bus 4 through 4KV breaker 52124. 4KV Buses 2 and 3 can be supplied from the SUT via 4KV breakers 52112 and 52/17 respectively. 4KV Buses 1 & 2 are normally tied together via tiebreaker 52/10. 4KV Buses 3 & 4 can be connected via tiebreaker 52119. During normal power operation 4KV Buses 1, 2, 4 and 5 are supplied via :UAT, and 4KV Bus 3 is supplied via the SUT. The 480V system is supplied from the 4160V system via 3-phase, 4160-480Y/277V SSTs. 4160V Bus No.1 supplies 2B SST, via feeder breaker 52/4. 4160V Bus No.2 supplies 2A & 2F SSTs, via feeder breaker 52113, fused primary switch 2A, and nonfused primary switch 2F. 4160V Bus No.3 supplies 2C & 2G SSTs, via feeder breaker 52/15, fused primary switch 2C, and nonfused primary switch 2G. 4160V Bus No.4 supplies 2D SST, via feeder breaker 52128. 4160V Bus No.5 supplies 2E SST, via feeder breaker 52/32. 3.7.1 4KV Bus 1, 2, 3 and 4 Breakers: KVAC Westinghouse, Air Circuit Breakers Type 50 DH-350E (horizontal drawout) Three-pole, electrically operated Interrupting Rating Continuous current @ 60cycles Rated voltage Max design voltage Interrupting current @ rated volt. Maximum interrupting current 350MVA 1200 ampere with Solenoid Operating Mechanism, & 3000 ampere with Motor Operated Stored Energy tSpring) Closing Mechanism

4. 16KV 4.76KV 48,000 Amperes 50,000 Amperes Page 13 of39 Revision 14 INFORMATION USE ONLY SD-002 SAFETY INJECTION SYSTEM A SI area cooling fan (HVH-6A or 6B, See SD-036 HV AC Systems) will start if a SI Pump or Containment Spray (CS) Pump starts. When one of the room HVH units is out of service the respective SI and CS pumps are considered inoperable.

However it is not necessary to rack out the affected pump breakers as only one HVH room cooler is required to support both SI pump motors and both CS pump motors.(ESR 95-00928) Electrical Power Supplies are: ---"A" SI Pump -480V Bus El "B" SI Pump -480V Bus El or E2 (Depending on which pump it is replacing) ,--"C" SI Pump -480V Bus E2 NOTE: The following starting duty limitations apply to the SI Pump. 1) IF the pump has not been run in the last hour, THEN 3 consecutive starts are allowed. 2) IF the pump has been started 3 times in the last hour AND neither of the last 2 starts was a run of at least 15 minutes, THEN no further starts are allowed for one hour. a. Any run in the previous hour is considered one of the 3 allowed starts. 3) IF the pump was run at least 15 minutes and stopped, THEN 2 consecutive starts are allowed with no waiting period. The function of the SI pumps is to complete the refill of the reactor vessel and ultimately return the core to a subcooled state. The flow from one SI pump and one RHR pump is sufficient to complete this refill function. Moreover, there is sufficient excess water delivered by the accumulators to tolerate a delay in starting the pumps. The starting sequence of the pumps and related emergency power equipment is designed so that delivery of the full rated flow is reached within 26 seconds after the process parameters reach the setpoints for the injection signal. This includes the time needed for the pump to run up to full rated delivery. 3.3 Boron Injection Tank (BIT) (No Longer Used as a Boron Injection Tank) Number/Type lIVertical Volume, tank 900 gallons Design pressure 1750 psig Design temperature 300°F Fluid Borated Water Material Austenitic stainless steel SI Page 14 of 43 Revision 14 INFORMATION USE ONLY 2F ELECTRICAL DISTRIBUTION KVAC-FIGURE-2 TO 115 KVSPAN BUS TO 220KV SWITCHYARO 1 MAIN I) 12 BUS 2 BUS 1 BUS" ,)20 BUS 5 j t:Y"'i--......... I .. I-- 2E 130B /37B 1)2B )8B )9B ) ) ) n US 1 BUS2AI BUS51 rd iii) 2G I) i) If I ;" '" fll.1c1 1 r.h [iJ 2 3 20 If-l ))) 15 17 pp-I"\-jMcCI 1 I 22 19 ) ) 1"\:4 6 E-1 E-2 i i i ) ) ?22B ?29B ) KIRK-KEY BREAKERS ) MCC-5 ) ) rllrtJ Iifc, "5) i1D 7 8 9 18 28B ?32A OS BUS ) KVACF02 HLC-08 NRC Written Exam 32. Given the following: -The plant is in MODE 3 following a load rejection and Reactor Trip from 100% RTP. -During the load rejection/trip event, a Pressurizer PORV opened for several seconds and reseated. -Pressurizer Relief Tank Pressure, Temperature and Level have increased: -Pressure = 4.8 PSIG. -Temperature = 153 OF (APP-003-B3, PRT HI TEMP is illuminated.) -Level = 80% -The CRSS directs implementation of OP-103, PRESSURIZER RELIEF TANK CONTROL SYSTEM to reduce PRT temperature. Which ONE (1) of the following methods will be used to reduce PRT temperature? A. Continuously fill the PRT with Primary Water and drain the PRT to the RCDT as necessary to maintain level between 70%-80%. B. Drain the PRT to the RCDT to 70%, then alternately refill with Primary Water and drain to RCDT. C. Fill the PRT with Primary Water to 90%, then alternately drain the PRT to the RCDT and refill with Primary Water. D. Monitor Automatic Primary Water spray, drain PRT to RCDT as necessary to maintain level between 70%-80%. 32 HLC-08 NRC Written Exam 32. 007 K4.01 OOllPRT/QUENCH TANK/2/112.6/2.9/RO/HIGHlNIAlNEW -20081PZR-008 Given the following: -The plant is in MODE 3 following a load rejection and Reactor Trip from 100% RTP. -During the load rejection/trip event, a Pressurizer PORV opened for several seconds and reseated. -Pressurizer Relief Tank Pressure, Temperature and Level have increased: -Pressure = 4.8 PSIG. -Temperature = 153 of (APP-003-B3, PRT HI TEMP is illuminated.) -Level = 80% -The CRSS directs implementation of OP-103, PRESSURIZER RELIEF TANK CONTROL SYSTEM to reduce PRT temperature. Which ONE (1) of the following methods will be used to reduce PRT temperature? A. Continuously fill the PRT with Primary Water and drain the PRT to the RCDT as necessary to maintain level between 70%-80%. 8:1 Drain the PRT to the RCDT to 70%, then alternately refill with Primary Water and drain to RCDT. C. Fill the PRT with Primary Water to 90%, then alternately drain the PRT to the RCDT and refill with Primary Water. D. Monitor Automatic Primary Water spray, drain PRT to RCDT as necessary to maintain level between 70%-80%. The correct answer is B. A: Incorrect -A continuous drain and fill is incorrect because the tank will fill faster than the drain capability. NO procedures support this action. B: Correct -For high PRT temperature, OP-103 directs draining the PRT to 70%-74%, then to fill with PW to cool the PRT. C: Incorrect -For rapid cooling and if level allows, the PRT is filled first then allowed to 'soak' for 10 minutes and then drained. Precautions and Limitations state that the tank should NOT be filled above 80% level. D: Incorrect -RNP has NO automatic Primary Water Spray available to the PRT. Tuesday, June 17,2008 1 :21 :08 PM 36 HLC-08 NRC Written Exam Exam Question Number: 32

Reference:

OP-103, Pages 6-10; APP-003-83, SO-059 PZRlPRT, Figure 1. KA Statement: Knowledge of PRTS design feature(s) and/or interlock(s) which provide for the following: Quench tank cooling. History: New -Written for HLC-08 NRC exam. Tuesday, June 17,20081:21:09 PM 37 8.0 INSTRUCTIONS 8.1 STARTUP Refer to GP-001. 8.2 NORMAL OPERATION REFERENCE USE Section 8.2.1 Page 1 of 3 8.2.1 Draining the PRT When Pressurizer Temperature is Greater Than Or Equal to 200°F ____ 1:...:.._---..:V.:::.e:..:..ri!y the following initial NOTE: PRT temperatures of greater than 120°F should be reduced by alternately adding Primary Water to the PRT and draining the PRT. (SER 93-007) Maximum cooling effect can be achieved by leaving the added Primary Water in the )"'" PRT for at least 10 minutes prior to draining. (SER 93-007) ", ' a. All the Prerequisites of Section 4.0 are complete.

b. The Pressurizer temperature is greater than or equal to 200°F AND PRT level is above 70%. c. Primary Water addition to the PRT is NOT in progress. (SER 93-007) NOTE: The following step is a continuous action step and should be performed when conditions are met. IOP-103 2. IF the PRT temperature is greater than or equal to 160°F, THEN Go To Step 8.2.1.6. Rev. 16 Page 6 of 23\

REFERENCE USE Section 8.2.1 Page 2 of 3 NOTE: Placing RC-523, PRT DRAIN, control switch in OPEN also opens LCV-10038, RCDT PUMP "8" SUCTION, and starts REACTOR COOLANT DRAIN TANK PUMP "8" if the control switches are in AUTO. 3. IF the normal drain path via the RCDT is available, AND a Containment Phase A Isolation signal is NOT present, THEN perform the following:

a. Open RC-523, PRT DRAIN. b. WHEN PRT level returns to between 70% and 74%, THEN close RC-523. NOTE: Instrument Air is loss to containment during a Containment Phase A Isolation signal unless PCV-1716 has been RESET OR placed in OVERRIDE.

IOP-103 4. IF the normal drain path via the RCDT is NOT available, OR a Containment Phase A Isolation signal is present, THEN perform the following when Instrument Air has been restored to the valves below: a. Open WD-1708, RCDT DRAIN TO CV SUMP. b. Place LCV-10038, RCDT PUMP "8" SUCTION, to the CLOSE position.

c. Open RC-523, PRT DRAIN. d. WHEN PRT level returns to between 70% and 74%, THEN perform the following:

w 1) Close RC-523. 2) Close WD-1708. 3) Place LCV-10038, RCDT PUMP "8" SUCTION, to the AUTO position 5. IF PRT temperature is greater than 120°F, THEN add Primary Water to the PRT in accordance with this procedure. Rev. 16 Page 7 of 231 " . .. REFERENCE USE CAUTION Section 8.2.1 Page 3 of 3 Pumping PRT contents through Containment Isolation valves following a PRT temperature increase in excess of 160°F can cause failure of the valves to seat due to debris accumulation from a failed PRT liner. (ESR 96-00608) IOP-103 6. IF the PRT temperature has been greater than or equal 160°F AND the liner has NOT been satisfactorily evaluated, THEN perform the following:

a. Verify CV Sump equipment aligned as follows: 1) CV Sump Pump breakers OPEN. -CV SUMP PUMP "A" on MCC 2 in CMPT 3M -CV SUMP PUMP "8" on MCC 1 in CMPT 5H 2) CV Sump Pump Discharge valves CLOSED -WD-1728, CONTAINMENT SUMP PUMP DISCHARGE AUTO ISOLATION

-WD-1723, CONTAINMENT SUMP PUMP DISCHARGE AUTO ISOLATION

b. Open WD-1708, RCDT DRAIN TO CV SUMP. c. Place LCV-10038, RCDT PUMP "8" SUCTION, to the CLOSE position.

d. Open RC-523, PRT DRAIN. e. WHEN PRT level returns to between 70% and 74%, THEN perform the following:

1) Close RC-523. 2) Close WD-1708. 3) Place LCV-10038, RCDT PUMP "8" SUCTION, to the AUTO position f. Place a Caution Tag on RC-523 switch that reads "If the PRT requires draining it shall be aligned to the CV sump only", This caution will remain in effect until the PRT internal coating evaluation is complete.

g. Consult RESS for a PRT internal coating evaluation.

Rev. 16 Page 8 of 23/ CONTINUOUS USE Section 8.2.2 Page 1 of 2 CHK (v') 8.2.2 Adding Primary Water to the PRT 1. Verify the following initial conditions are satisfied: NOTE: PRT temperatures of greater than 120°F should be reduced by alternately adding Primary Water to the PRT and draining the PRT. (SER 93-007) Maximum cooling effect can be achieved by leaving the added Primary Water in the PRT for at least 10 minutes prior to draining. (SER 93-007) a. b. All the Prerequisites of Section 4.0 are complete. Pressurizer Relief Tank level is less than 80%. Draining the PRT is NOT in progress. (SER 93-007) CAUTION Operating two Primary Water Pumps can cause inadvertent filling of the RCS if the RCS is depressurized and vented through a PORV when two Primary Water Pumps are operating. Water may makeup to the RCS via the PRT spargers if the PRT is filled faster than it can vent. IOP-103 2. Verify a Primary Water Pump is OPERATING.

3. Open RC-519A & 8, PW TO CV ISO. 4. Open RC-519C, PW TO PRT ISO. 5. Monitor PRT level for an increase.

6. IF the expected increase does NOT occur, THEN stop filling AND investigate.

Rev. 16 Page 9 of 231 CONTINUOUS USE Section 8.2.2 Page 2 of 2 CHK (v') NOTE: Increasing PRT level will cause PRT pressure to increase, possibly to the high pressure alarm setpoint of 5 psig. IOP-103 7. WHEN PRT level is between 70% and 80%, THEN perform the following: .j a. Stop the Primary Water Pump. b. Close RC-519A & B. c. Close RC-519C. 8. IF PRT level is greater than or equal to 83% OR PRT temperature is greater than 120°F, THEN lower the PRT level in accordance with this procedure.

9. IF RC-519A or RC-519B do not fully close, THEN relieve the hydraulic lock in accordance with this procedure.

Rev. 16 Page 10 of 231 ALARM PRT HI TEMP AUTOMATIC ACTIONS 1. Not Applicable CAUSE 1. Opening of Pressurizer Safety or PORV 2. Pressurizer Safety or PORV leakage OBSERVATIONS

1. PRT Level (LI-470) 2. PRT Pressure (PI-472) 3. PRT Temperature (TI-471) 4. Pressurizer Safety Valve Line Temperatures (TI-465, TI-467, TI-469) 5. PORV Discharge Line Temperature (TI-463) ACTIONS 1. Alternately add Primary Water to the PRT and drain the PRT using OP-103. DEVICE/SETPOINTS

1. TC-471 I 150°F POSSIBLE PLANT EFFECTS 1. None Applicable REFERENCES

1. CWD B-190628, Sheet 461 , Cable L 2. OP-103, Pressurizer Relief Tank Control System I APP-003 Rev. 37 APP-003-B3 Page 17 of 531 IVSW RC-516 I RC-533 Gas Analyzer Vent.. "'4 Header SYSTEM SIMPLIFIED DIAGRAM PZR-FIGURE-l PZR fj"---Primary Water RC-519A RC-5198 RC-523 Ipzrf011 INFORMATION USE ONLY Loop c Hot Leg B HLC-08 NRC Written Exam 33. Given the following:

-The plant is in EPP-10, LONG TERM RECIRCULATION. -APP-002-ES, SI PMP COOL WTR LO FLOW alarm is received. Which ONE (1) of the following describes the cause and effect of the alarm? A. Loss of CCW to the SI Pump Bearing Heat Exchangers; can result in overheating the pump bearings. B. Loss of CCW to the SI Pump Seal Coolers; can result in loss of SI Pump seals. C. Loss of SW to the SI Pump Bearing Heat Exchangers; can result in overheating the pump bearings. D. Loss of SW to the SI Pump Seal Coolers; can result in loss of SI Pump seals. 33 HLC-08 NRC Written Exam 33.008 A4.05 001lCCW/2/112.7/2.5IROILOWINIAINEW -2008/CCW-009 Given the following: -The plant is in EPP-10, LONG TERM RECIRCULATION. -APP-002-E5, SI PMP COOL WTR LO FLOW alarm is received. Which ONE (1) of the following describes the cause and effect of the alarm? A. Loss of CCW to the SI Pump Bearing Heat Exchangers; can result in overheating the pump bearings. B:' Loss of CCW to the SI Pump Seal Coolers; can result in loss of SI Pump seals. C. Loss of SW to the SI Pump Bearing Heat Exchangers; can result in overheating the pump bearings. D. Loss of SW to the SI Pump Seal Coolers; can result in loss of SI Pump seals. The correct answer is B. A: Incorrect -CCW supplies SI Pump Seal Coolers, NOT Bearing Heat Exchangers. B: Correct -CCW Low flow (50 GPM) to the Seal Coolers initiates the alarm. C: Incorrect -SW does NOT supply SI Pump Bearing Heat Exchangers. SW does supply SI Pump Thrust Bearing cooling bath, but has NO alarm. D: Incorrect -SW does NOT supply SI Pump Seal Coolers. SW does supply SI Pump Thrust Bearing cooling bath, but has NO alarm. Exam Question Number: 33

Reference:

APP-002-E5; SD-013 CCW, Page 13. KA Statement: Ability to manually operate and/or monitor in the control room: Normal CCW-header total flow rate and the flow rates to the components cooled by the CCWS. History: New -Written for HLC-08 NRC exam. Tuesday, June 17,2008 1 :21 :09 PM 38 ALARM SI PMP COOL WTR LO FLOW AUTOMATIC ACTIONS 1 . None Applicable CAUSE 1. Misaligned valve or leak in lines supplying CCW to or from the SI Pumps. OBSERVATIONS

1. None applicable (see actions) ACTIONS APP-002-E5 1 . IF long term post accident recirculation is NOT in progress, THEN dispatch personnel to check CCW flow to the SI Pump Seal Coolers, FIC-658. 2. IF SI Pumps are operating under non-emergency conditions OR long term recirculation, THEN stop the pumps. 3. IF a loss of CCW has occurred, THEN refer to AOP-014. 4. IF valve alignment is NOT correct, THEN verify correct valve alignment using OP-306. 5. IF a CCW leak is present, THEN isolate the leak. DEVICE/SETPOINTS

1. FIC-658 I 50 gpm

-Loss of SI Pump Seals. (If in recirculation mode .. ____ .-J REFERENCES

1. AOP-014, Component Cooling Water System Malfunction

2. ITS LCO 3.5.2, 3.5.3 3. Flow Drawing, 5379-376, Sh 4 4. CWD B-190628, 488G 5. OP-306, Component Cooling Water System I APP-002 Rev. 54 Page 55 of 651 SD-013 COMPONENT COOLING WATER SYSTEM used. When the evaporators were in service, CCW was supplied to the concentrator section of the evaporator to condense the steam that is boiled off in the concentrator.

The distillate was further cooled in the condensate cooler before it was pumped to the monitor tanks (MT). The CCW flow to the concentrator was throttled for proper operation of the evaporator. 3.6.7 Sample Heat Exchangers CCW supplies cooling water for the following sample heat exchangers:

1. Reactor Coolant 2. Pressurizer Liquid Space 3. Pressurizer Steam Space 4. Steam Generator Blowdown (one for each A, B, & C) 5. Post Accident Sampling System Each cooler can be isolated by manually operated valves. 3.6.8 Waste Gas Compressor CCW is' supplied to these compressors for the purpose of cooling the water used for sealing the shaft seals on either side of the compressor and also for makeup to the separator tank automatically when water is needed. The gas compressor has to be running before automatic makeup will function.

3.6.9 Residual Heat Removal Pumps CCW is supplied to the sealing water heat exchanger on these pumps. Emergency cooling connections are also available so these pumps may be operated using an external cooling source if CCW ever has to be isolated from these components. CCW is supplied to on each of the 3 SI pumps. Emergency cooling connections are also available so these pumps may be operated using an external cooling source if CCW ever has to be isolated from these components. 3.6.11 Containment Spray (CS) Pumps CCW is supplied to the seal water heat exchanger on each of the CS pumps to cool the seal water being used to seal and cool the mechanical seal on the pump shaft. 3.6.12 Spent Fuel Pit Heat Exchanger ccw Page 13 of38 Revision 9 INFORMATION USE ONLY HLC-08 NRC Written Exam 34. Given the following: -The plant is in MODE 3. -The crew is performing a plant heatup. -Pressurizer level is 50%. -BOTH Pressurizer Liquid Space and Steam Space temperature indicators read 622 of. Which ONE (1) of the pressure indicators most accurately represents current Pressurizer Pressure? A. PI-444 = 1783 PSIG. B. PI-445 = 1798 PSIG. C. PI-455 = 1813 PSIG. D. PI-456 = 1828 PSIG. 34 HLC-08 NRC Written Exam. 34. 010 KS.01 001IPZR PRESSURE CONTROU2/1/3.S/4.0IROIHIGH/N/A/NEW -2008ITHERMO CHAP 3-008 Given the following: -The plant is in MODE 3. -The crew is performing a plant heatup. -Pressurizer level is 50%. -BOTH Pressurizer Liquid Space and Steam Space temperature indicators read 622 of. Which ONE (1) of the pressure indicators most accurately represents current Pressurizer Pressure? A. PI-444 = 1783 PSIG. 8:0' PI-445 = 1798 PSIG. C. PI-455 = 1813 PSIG. D. PI-456 = 1828 PSIG. The correct answer is B. A: Incorrect -Subtracted 14.7 PSI from corrected answer. B: Correct -Using Steam Tables, interpolate for 622 of. 620 of = 1786 .. 9 PSIA, 624 0 F= 1839.0 PSIA. Therefore 622 of = 1812.95 PSIA. Convert 1812.95 PSIA to PSIG: 1812.95 PSIA -14.7 PSI = 1798.25 PSIG C: Incorrect -Listed answer without adjusting for PSIA. D: Incorrect -Added 14.7 PSI instead of subtracting to obtain PSIG. Exam Question Number: 34

Reference:

Steam Tables; SD-059 PZRlPRT, Figure 1. KA Statement: Knowledge of the operational implications of the following concepts as the apply to the PZR PCS: Determination of condition of fluid in PZR, using steam tables. History: New -Written for HLC-08 NRC exam. Tuesday, June 17, 20081:21:09 PM 39 Gas Analyzer IVSW Vent.. 111'4 Header SYSTEM SIMPLIFIED DIAGRAM PZR-FIGURE-I PZR Primary ... J -'. fj'lo--Water ) .. .. RC-523 RC-519A RC-5198 lpzrf01 I INFORMATION USE ONLY loop c Hot leg 8 HLC-08 NRC Written Exam 35. Given the following: -The plant is operating at 6% RTP preparing for Turbine roll. -PZR level channel LT-459 failed 4 hours ago. The bistables have been tripped and all actions are complete per AOP-025, RTGS INSTRUMENT FAILURE. -PZR level is currently 25% on Channels LT-460 and 461. Which ONE (1) of the following describes the effects on the plant if PZR level channel L T-461 fails HIGH? A. A Reactor Trip due to high PZR level. B. A Reactor Trip due to low PZR level. C. The Reactor will NOT trip, signal is blocked by P-7. D. The Reactor will NOT trip, signal is blocked by P-1 O. 35 HLC-08 NRC Written Exam 35. 012 K6.01 001/RX PROTECTION/2/1/2.8/3.3IROIHIGH/NINCOOK -20021RPS-006 Given the following: -The plant is operating at 6% RTP preparing for Turbine roll. -PZR level channel L T -459 failed 4 hours ago. The bistables have been tripped and all actions are complete per AOP-025, RTGB INSTRUMENT FAILURE. -PZR level is currently 25% on Channels LT-460 and 461. Which ONE (1) of the following describes the effects on the plant if PZR level channel LT-461 fails HIGH? A. A Reactor Trip due to high PZR level. B. A Reactor Trip due to low PZR level. Cr The Reactor will NOT trip, signal is blocked by P-7. D. The Reactor wi" NOT trip, signal is blocked by P-10. The correct answer is C. A: Incorrect -A High Level trip signal is generated from the channel failure, but it is blocked by P-7. B: Incorrect -A trip will NOT occur due to low level. C: Correct -With Channel LT-459 in the tripped condition the High level Rx Trip Signal will be made up for 1 channel (1/2 coincidence on remaining channels). The level control selector switch for the Pressurizer is in the 461 REPLACE 459 position with channel LT-461 as the controlling channel. When it fails high a High Level Rx Trip signal is generated but it is blocked by P-7 (Reactor and Turbine power both below 10%). 0: Incorrect -Signal is blocked by Permissive P-7, (Turbine Power) NOT P-10, (NI Power). Exam Question Number: 35

Reference:

SO-011 RPS, Pages 20-21, 29, Figure 31. KA Statement: Knowledge of the effect of a loss or malfunction of the following will have on the RPS: Bistables and bistable test equipment. History: Modified; removed LCO actions. Tuesday, June 17, 2008 1 :21 :09 PM 40 SD-011 4.1.5.10 a. b. 4.1.5.11 a. b. 4.1.5.12 a. b. ,'/4.1.5.13

a. RPS REACTOR PROTECTION SYSTEM RCP Bus Underfrequency Trip (Not a direct reactor trip) (Figure 28) The RCP Bus Underfrequency Trip trips the RCP breakers.

RCP breaker open signal trips the reactor which provides protection for the Reactor against DNB as a result of an underfrequency on more than one RCP Bus. This trip occurs when an underfrequency condition exists on 2 out of 3 RCP Buses when above 10% (P-7). This trip is automatically blocked below 10% (P-7). This trip assures a Reactor Trip Signal is generated before the Low Flow Trip Setpoint is reached. Setpoint -58.2 Hertz High PZR Pressure Trip (Figure 29) The High PZR Pressure Trip provides protection for the Reactor Coolant System (RCS) against over pressurization and limits the range of required protection from the OTilT Trip. This trip occurs when 2 out of 3 PZR Pressure Signals exceed the trip setpoint. Setpoint -PC-455A, PC-456A, PC-457A/2376 psig Low PZR Pressure Trip (Figure 30) The Low PZR Pressure Trip provides protection against excessive void formation in the Reactor which could lead to a DNB ratio (DNBR) of < 1.17 and limits the necessary range of protection afforded by the OT il T . This trip occurs when 2 out of 3 PZR Pressure Signals decreases below the trip setpoint. This trip is automatically blocked below 10% (P-7). This trip is dynamically compensated based on the rate of change in pressure. Setpoint -PC-455C, PC-456C, PC-457C PM-455A, PM-456A, PM-457 A Trip Setpoint Lead Time Constant Lag Time Constant 1844 psig 10 sec. 1 sec. High Pressurizer (PZR) Water Level Trip (Figure 31) The High PZR Water Level Trip provides a back-up to the High PZR Pressure Page 20 of30 Revision 8 INFORMATION USE ONLY

b. 4.1.5.14 a. b. 4.1.5.15 a. b. 4.1.5.16 a. b. RPS REACTOR PROTECTION SYSTEM Trip and prevents the PZR Safety and Relief Valves from relieving water for credible accident conditions.

This trip occurs when 2 out of 3 PZR Water Level Signals exceeds the trip setpoint. This trip is automatically blocked below 10% (P-7). Setpoint -LC-459A, LC-460A, LC-461A/91 % of span SteamlFeedwater Flow Mismatch Trip (Figure 32) The Steam/Feedwater Flow Mismatch Trip provides protection for the Reactor against an anticipated Loss of Heat Sink. This trip occurs when lout of 2 flow elements sense that Feedwater Flow is < Steam Flow and lout of 2 Steam Generator(S/G) Level Elements decrease below the setpoint in any S/G. Setpoint -FC"7478A, FC-478B/O.64 x 10 6 lbs/Hr FC-488A, FC-488B/O.64 x 10 6 lbs/Hr FC-498A, FC-498B/O.64 x 10 6 lbs/Hr AND LC-474B, LC-475B/30% of Span LC-484B, LC-485B/30% of Span LC-494B, LC-495B/30% of Span S/G Low-Low Water Level Trip (Figure 33) The S/G Low-Low Water Level Trip provides protection for the Reactor by preventing operation without adequate heat removal capability in the event of a sustained Steam/Feedwater Flow mismatch which is sufficiently small not to be sensed by the Steam/Feedwater Flow Mismatch Trip. This trip occurs when 2 out of 3 S/Gs Narrow Range Level Elements on lout of 3 S/Gs decrease below the setpoint. Setpoint -LC-474A, LC-475A, LC-476A/16% of Span LC-484A, LC-485A, LC-486A/16% of Span LC-494A, LC-495A, LC-496A/16% of Span Safeguards Signal Trip The Engineered Safeguards Signal Trips ensure that the Reactor will be shut down during a severe accident. This trip is initiated if the Engineered Safety Features Actuation System is automatically or manually actuated. Setpoint -Refer to SD-006, Engineered Safety Features System Page 21 of 30 Revision 8 INFORMATION USE ONLY SD-011 REACTOR PROTECTION SYSTEM ATTACHMENT 10.1 Page 2 of3 REACTOR PROTECTION SYSTEM PERMISSIVES PERMISSIVE NUMBER DERIVATION FUNCTION P-7 2/4 Power Ranges above Enables the following trips: setpoint (10% from P-lO) 1. RCS Low Flow OR 2. RCP Breakers Open 112 Turbine First Stage 3. UV Pressure above setpoint 4. Turbine Trip (10%) 5. PZR Low Pressure 6. PZR High Level 3/4 Power Ranges below Blocks the following reactor setpoint (10%) from P-lO trips: AND 1. RCS Low Flow 2/2 Turbine First Stage 2. RCP Breakers Open Pressure below setpoint 3. UV (10%) 4. Turbine Trip 5.PZR Low Pressure 6.PZR High Level P-8 2/4 Power Ranges above Enables Reactor Trip on low setpoint (40 %) flow in a single loop 3/4 Power Ranges below Blocks Reactor Trip on low setpoint (40 %) flow in a single loop RPS Page 29 0[30 Revision 8 INFORMATION USE ONLY PRESSURIZER HIGH WATER LEVEL REACTOR TRIP LOGIC RPS-FIGURE-31 I PRESSURIZER HIGH WATER LEVEL REACTOR TRIP LOGIC PRESSURIZER LEVEL INSTRUMENTS II 2/3 AND REACTOR TRIP III P-7 I rpsf31 I au ONS REPORT for CookExam200.?-J 1.045 001lSRO/045IINPO -DIRECT/53141bob028-'2:.'Z:22lRO#"NNSRO# 04113H Given the following plant conditions: -The plant is operating at 6% power preparing for Turbine roll. -PRZ level channel (1) NLP-151 failed 4 hours ago. The bistables have been tripped and all actions are complete as per 01-0HP-4022-013-01 0, Pressurizer Level Instrument Malfunction. -PRZ level is currently 25% on channel (2) NLP-152 and (3) NLP-153. Which ONE of the following describes the effects on the plant if PRZ level channel (3) NLP-153 fails high? Assume NO operator actions. A. A Reactor Trip due to high pressurizer level. B. A Reactor Trip and Safety injection due to a loss of pressurizer level. C. The Reactor will not trip, continue with the plant startup. The Reactor will not trip, however the plant must be placed in hot shutdown. ANSWER: D -With Channel 1 NLP-151 in the tripped condition the High level Rx Trip signal will be made up for 1 channel (1/2 coincidence on remaining channels). The level control selector switch for the pressurizer is in the 2/3 position with channel 3 NLP-153 as the controlling channel. When it fails high a High Level Rx Trip signal is generated but it is blocked by P-7 (Reactor and Turbine power both below 10%). Plant Shutdown is required due to Tech Spec 3.3.1 (3.0.3). A -Incorrect -A High Level trip signal is generated from the channel failure (and an actual high level would occur -charging lowers, letdown isolates, pressurizer fills from seal injection) but it is blocked by P-7. B -Incorrect -A trip & SI will not occur due to loss of level. C -Incorrect -Startup can not continue due to Tech Specs and a trip if power is increased to > 10%. Lesson Plan/Obj: RO-C-00202 / #27

Reference:

SOD-00202-003, Pressurizer Level Control Pressurizer (PRZ) Level Control Malfunction -Equipment Control -Knowledge of limiting conditions for operations and safety limits. Category 1: SRO Category 2: 045 Category 3: INFO -DIRECT Category 4: 5314 Category 5: 000028 -2.2.22 Category 6: RO#NA Category 7: SRO#041 Category 8: 3H Tuesday, June 10,20087:24:07 AM 1 HLC-08 NRC Written Exam 36. Given the following: -The plant is operating at 100% RTP. -ITS 3.3.2 requires all channels of ESFAS instrumentation to be OPERABLE. -Surveillance testing of ESFAS Train "A" is scheduled to commence in 1 hour. What specific provision of ITS or facility license allows for the scheduled testing of an ESFAS train without entry into the associated Conditions and Required Actions? A. The Facility License specifically exempts Surveillance Testing from entry into ITS LCOs. LCO entry is NOT required if the COMPLETION TIME is met. B. SR 3.0.2 allows surveillances to be performed within 1.25 times the specified interval. LCO entry is NOT required. C. ITS 3.3.2, SURVEILLANCE REQUIREMENTS, allows entry into associated Conditions and Required Actions to be delayed up to 6 hours, provided the redundant Train is OPERABLE. D. ESFAS Train testing places the actuation bistables in the TRIPPED condition, which provides the Safeguards Function. LCO entry is NOT required. 36 HLC-08 NRC Written Exam 36. 013 G2.2.38 001IESFAS/2/1/3.6/4.5IROIHIGHIN/A/NEW

• 2008IESF*008 Given the following:

-The plant is operating at 100% RTP. -ITS 3.3.2 requires all channels of ESFAS instrumentation to be OPERABLE. -Surveillance testing of ESFAS Train "A" is scheduled to commence in 1 hour. What specific provision of ITS or facility license allows for the scheduled testing of an ESFAS train without entry into the associated Conditions and Required Actions? A. The Facility License specifically exempts Surveillance Testing from entry into ITS LCOs. LCO entry is NOT required if the COMPLETION TIME is met. B. SR 3.0.2 allows surveillances to be performed within 1.25 times the specified interval. LCO entry is NOT required. ITS 3.3.2, SURVEILLANCE REQUIREMENTS, allows entry into associated Conditions and Required Actions to be delayed up to 6 hours, provided the redundant Train is OPERABLE. D. ESFAS Train testing places the actuation bistables in the TRIPPED condition, which provides the Safeguards Function. LCO entry is NOT required. The correct answer is C. A: Incorrect -No exception for entry into LCO REQUIRED ACTION simply for surveillance testing exists in the Facility License. B: Incorrect -SR 3.0.2 does allow for surveillances to be performed within 1.25 times the specified interval, but the specification is applied for completion times only, NOT to determine if LCO entry is required. C: Correct -ITS 3.3.2, SURVEILLANCES, NOTE 2 states: "When a channel or train is placed in an inoperable status solely for the performance of required SURVEILLANCES, entry into the associated Conditions and Required Actions may be delayed for up to 6 hours provided the redundant train is OPERABLE." D: Incorrect -Safeguards devices may be considered OPERABLE if placed in their safeguards actuate position (e. g. closing an inoperable isolation valve would allow the valve's safeguards function to be considered met), but the function bistables for the Train in test are tripped (and untripped when complete) individually. Exam Question Number: 36

Reference:

ITS 3.3.2, Pages 3.3-20 and 3.3-24; SR 3.0.2, Page 3.0-4. KA Statement: Knowledge of conditions and limitations in the facility license. History: New -Written for HLC-08 NRC exam. Tuesday, June 17, 2008 1 :21:10 PM 41 .--. 3.3 INSTRUMENTATION ESFAS Instrumentation 3.3.2 3.3.2 Engineered Safety Feature Actuation System (ESFAS) Instrumentation LCO 3.3.2 The ESFAS instrumentation for each Function in Table 3.3.2-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.2-1. ACTIONS ------. ---- .. -------.. -----------. --. NOTE -. --. --.. -.... -. ---..... -... -. ------. Separate Condition entry is allowed for each Function . ........ --.................................................................... . CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions A.I Enter the Condition Immediately with one or more referenced in required channels or Table 3.3.2-1 for the trains inoperable. channel(s) or train(s). B. One channel or train B.1 Restore channel or 48 hours inoperable. train to OPERABLE status. OR B.2.1 Be in MODE 3. 54 hours AND B.2.2 Be in MODE 5. 84 hours (continued) HBRSEP Unit No. 2 3.3-20 Amendment No. 176 .-ESFAS Instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS ...... -... ---. -. --- .............. ---. NOTES* .... -.... -.......... -....... -.. -.. . 1. Refer to Table 3.3.2*1 to determine which SRs apply for each ESFAS Function.

2. When a channel or train is placed in an inoperable status solely for the performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the redundant train is OPERABLE . ...........

-_ .............. -.................................... -....... -...... -SURVEILLANCE FREQUENCY SR 3.3.2.1 Perform CHANNEL CHECK. 12 hours SR 3.3.2.2 Perform ACTUATION LOGIC TEST. 31 days on a STAGGERED TEST BASIS SR 3.3.2.3 Perform MASTER RELAY TEST. 18 months SR 3.3.2.4 Perform COT. 92 days SR 3.3.2.5 Perform SLAVE RELAY TEST. 18 months SR 3.3.2.6 ....*........ -....* NOTE* ................... Verification of setpoint not required for manual initiation functions . ....... -................... -........... __ ..... Perform TADOT. 18 months SR 3.3.2.7 Perform CHANNEL CALIBRATION. 18 months HBRSEP Unit No. 2 3.3*24 Amendment No. 176 SR Applicability 3.0 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SR 3.0.2 SR 3.0.3 HBRSEP Unit No. 2 SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Survei 11 ance or between performances of the Survei 11 ance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3. Surveillances do not have to be performed on inoperable equipment or variables outside specified limits. The speci fi ed Frequency for each SR is met if the Survei 11 ance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met. For Frequencies specified as "once," the above interval extension does not apply. If a Completion Time requires periodic performance on a "once per ... " basis, the above Frequency extension appl ies to each performance after the initial performance. Exceptions to this Specification are stated in the individual Specifications. If it is di scovered that a Survei 11 ance was not performed wi thi n its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours or up to the limit of the specified Frequency, whichever is less. This delay period is permitted to allow performance of the Surveillance. (continued) 3.0-4 Amendment No. 203 HLC-08 NRC Written Exam 37. Given the following: -The plant is operating at 100% RTP. -HVH-3 has been experiencing intermittent problems and the fan has been stopped. -I & C is troubleshooting the problem. -APP-002-E6, HVH-1/2/3/4 EMERGENCY CONTROL is illuminated. -CV RECIRC FAN HVH-3 indication on the RTGB is extinguished. -An inadvertent Safety Injection is received. Which ONE (1) of the following describes the current status of HVH-3? The Fan is ... A. stopped and will NOT start due to the LOCAL-REMOTE switch being in LOCAL. B. stopped, but can be started from the RTGB using the control switch. C. running because the SI signal is independent of the LOCAL-REMOTE switch. D. running because it started on Low Air Flow following completion of the SI sequencer. 37 HLC-08 NRC Written Exam 37.022 A3.01 OOllCTMT COOLING/2/1/4.1I4.3IROIHIGHlNINNEW -200S/CVHVAC-00S Given the following: -The plant is operating at 100% RTP. -HVH-3 has been experiencing intermittent problems and the fan has been stopped. -I & C is troubleshooting the problem. -APP-002-E6, HVH-1/2/3/4 EMERGENCY CONTROL is illuminated . .. CV RECIRC FAN HVH-3 indication on the RTGB is extinguished. -An inadvertent Safety Injection is received. Which ONE (1) of the following describes the current status of HVH-3? The Fan is ... A'I stopped and will NOT start due to the LOCAL-REMOTE switch being in LOCAL. B. stopped, but can be started from the RTGB using the control switch. C. running because the SI signal is independent of the LOCAL-REMOTE switch. D. running because it started on Low Air Flow following completion of the SI sequencer. The correct answer is A. A: Correct -The Local-Remote switch inhibits automatic starts of the HVH-3 fan when in LOCAL. B: Incorrect -In LOCAL, the only place that component can be operated is the Local Control Station. C: Incorrect -LOCAL-REMOTE Switch must be placed in REMOTE for ANY Auto signals to actuate equipment. D: Incorrect -Low Air flow will alarm (APP-002-CS), but provides NO Auto start feature of the Fan. Exam Question Number: 37

Reference:

APP-002-E6; APP-002-CS. KA Statement: Ability to monitor automatic operation of the CCS, including: Initiation of safeguards mode of operation. History: New -Written for HLC-08 NRC Exam. Tuesday, June 17,20081:21:10 PM 42 APP-002-E6 ALARM 1. HVH-1/2/3/4 EMERGENCY CONTROL AUTOMATIC ACTIONS 1 . None Applicable CAUSE 1. HVH-1, 2, 3, OR 4 Emergency Control Switches in LOCAL position. OBSERVATIONS

1. HVH Status Lights out on RTGB. 2. Local Switch position ACTIONS 1. Determine reason for Local Control. 2. IF switches have NOT been intentionally positioned, THEN return emergency control switches to normal. 3. WHEN the switch has been restored to normal, THEN reset the HVH Damper using OP-921. 4. IF HVH Units are required to be operable AND the alarm is illuminated, THEN monitor all CV HVH Cooler Status Lights on RTGB until the switch is repositioned.

DEVICE/SETPOINTS

1. None Applicable POSSIBLE PLANT EFFECTS Fan can NOT be operated from Control Room. Fans will NOT auto start on safety injection. (Inoperable)

3. Fan motor overload/trip alarm will NOT actuate on motor overload.

REFERENCES

1. ITS LCO 3.6.6 2. CWO B-190628, Sh 511,512,513,514,5150

3. OP-921, Containment Air Handling I APP-002 Rev. 54 Page 56 of 651 ALARM HVH-3 AIR FLOW LOST AUTOMATIC ACTIONS 1 . None Applicable CAUSE 1. Roughing filter clogged (Only installed during plant shutdown)

2. Mechanical failure fan/motor

3. Flow switch failure OBSERVATIONS

1. HVH Alarms for High Vibration

2. HVH Status Lights on RTGB ACTIONS APP-002-C5

1. IF the Hi Vibration alarm (APP-002-A?)

is also received for HVH-3, THEN stop HVH-3. 2. IF available, THEN start a standby Containment Recirc Cooler Fan. 3. Initiate actions to determine cause/validity of alarm 4. IF fan is inoperable, THEN refer to ITS LCO 3.6.6. DEVICEISETPOINTS

1. FS-4?15/ POSSIBLE PLANT EFFECTS 1. Possible entry into TECH SPEC LCO. 2. CV elevated temperature.

REFERENCES

1. ITS LCO 3.6.6 2. Flow Diagram G-190304, Sh 1 3. CWD B-190628, Sh 513 I APP-002 Rev. 54 Page 33 of 651 HLC-08 NRC Written Exam 38. Given the following: -A Reactor Trip and Safety Injection have occurred.

-The crew is performing actions of PATH-1. -CV Spray initiated on increasing Containment pressure. -CV Spray Pump "A" breaker has tripped. -CV Spray Pump "B" is operating normally. -All other ECCS equipment is operating as required. Which ONE (1) of the following is the impact on the plant and the action required? Accident analysis has determined that Containment pressure ... A. MAY exceed design. Immediately transition to FRP-J.1, RESPONSE TO HIGH CONTAINMENT PRESSURE, due to the RED condition. B. will NOT exceed design. Continue in PATH-1 until SPDS is reset and transition to another procedure is directed. C. MAY exceed design. Continue in PATH-1 until SPDS is reset and transition to another procedure is directed. D. will NOT exceed design. Transition to FRP-J.1 if CV Spray Pump "A" CANNOT be restarted. 38 HLC-OS NRC Written Exam 3S. 026 Al.OI OOIlCTMT SPRAY/2/113.9/4.2IROIHIGH/N/NRNP AUDIT BANK/CSS-005 Given the following: -A Reactor Trip and Safety Injection have occurred. -The crew is performing actions of PATH-1. -CV Spray initiated on increasing Containment pressure. -CV Spray Pump "A" breaker has tripped. -CV Spray Pump "8" is operating normally. -All other ECCS equipment is operating as required. Which ONE (1) of the following is the impact on the plant and the action required? Accident analysis has determined that Containment pressure ... A. MAY exceed design. Immediately transition to FRP-J.1, RESPONSE TO HIGH CONTAINMENT PRESSURE, due to the RED condition. 8:1 will NOT exceed design. Continue in PATH-1 until SPDS is reset and transition'to another procedure is directed. C. MAY exceed design. Continue in PATH-1 until SPDS is reset and transition to another procedure is directed. D. will NOT exceed design. Transition to FRP-J.1 if CV Spray Pump "A" CANNOT be restarted. The correct answer is 8. A: Incorrect -Minimum safeguards equipment (1 train of CV Spray and 2 HVH units) is required to maintain design parameters. CV Spray Pump "8" and all other ECCS equipment is operating. 8: Correct -Pressure will NOT exceed design if 1 train of CV spray and 2 HVH units are operating. The crew should continue in PATH-1 until told to transition to another procedure if applicable. C: Incorrect -Minimum safeguards equipment (1 train of CV Spray and 2 HVH units) is required to maintain design parameters. CV Spray Pump "8" and all other ECCS equipment is operating. 0: Incorrect -Transition to FRP-J.1 does NOT depend on the restart of CV Spray Pump "A". It depends on CV pressure exceeding 42 PSIG or NO CV Spray available above 10 PSIG. Tuesday, June 17,20081:21:10 PM 43 HLC-08 NRC Written Exam Exam Question Number: 38

Reference:

PATH-1 BD, Pages 7, 8 for grid location E-3 (E-1 AND E-2 energized); SD-024, Containment Spray, Page 7. KA Statement: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CSS controls including: Containment pressure. . History: Modified; removed operator actions to restart CV Spray. Tuesday, June 17, 20081:21:10 PM 44 GRID WOG BASIS/DIFFERENCES STEP E-2 D-2 2 2 RNPSTEP TURBINE TRIPPED WOG BASIS PURPOSE: To ensure that the turbine is tripped BASIS: The turbine is tripped to prevent an uncontrolled cooldown of the RCS due to steam flow that the turbine would require. RNP DIFFERENCES/REASONS There are essentially no differences. Interpretation The ERG contains a substep to check ALL Stop Valves closed. At RNP there are two stop valves, if both are closed the turbine is considered tripped. SSD DETERMINATION This is not an SSD. RNP STEP (RNO) TRIP OR RUN BACK TURBINE WOGBASIS See ERG step 2 above. RNP DIFFERENCES/REASONS The RNP step has added an additional action to run the Turbine back should the manual trip be unsuccessful. SSD DETERMINATION This is an SSD per criterion

10. RNP STEP E-1 AND E-2 ENERGIZED I PATH-1-BD Rev 18 Page 7 of 951 GRID WOG BASIS/DIFFERENCES

([;-3) STEP F-2 3 F-3 3 I PATH-1-BD WOG BASIS PURPOSE: To ensure electrical power to at least one emergency bus BASIS: AC power must be verified from either offsite sources or the diesel generators to ensure adequate power sources to operate the safeguards equipment. At least one train of safeguards equipment is required to deal with emergency conditions. If at least one train is not available, the operator should try to quickly restore one train, e.g., start a diesel generator and load it on the emergency bus. If at least one train cannot be restored quickly, the operator should transfer to ECA-O.O, LOSS OF ALL AC POWER. Guideline ECA-O.O is developed and structured to address the condition where all ac emergency power is lost. It is entered on the symptom of all ac emergency busses being deenergized. Its objective is to cope with the loss of ac emergency power until at least one ac emergency bus can be energized. ECA-O.O should not be entered if at least one ac emergency bus is energized since the other optimal recovery guidelines and function restoration guidelines contain guidance that accommodates multiple failures. They use available equipment to mitigate events whether plant systems are at full capacity, minimum safeguards capacity or degraded capacity. The availability of minimum safeguards capacity is not a requirement for being in the other optimal recovery guidelines and function restC?ration guidelines. For example, the core cooling function restoration guidelines provide guidance for the use of available equipment in degraded systems to mitigate inadequate core cooling (ICC) (e.g., ICC analyses show that only one high head SI pump is needed to prevent ICC even though one high head SI pump may not be sufficient to mitigate design basis transients within their design basis acceptance criteria). It is also desirable to have power to all ac emergency busses. If power is available to only one train, the operator should initiate attempts to restore power to the other train while continuing with the next step in the guideline to deal with the emergency condition. RNP DIFFERENCES/REASONS There are essentially no differences. SSD DETERMINATION This is not an SSD. RNP STEP EITHER E-1 OR E-2 ENERGIZED (with transition to EPP-1) WOGBASIS See step 3 above. RNP DIFFERENCES/REASONS There are essentially no differences. SSD DETERMINATION This is not an SSD. RNP STEP ATTEMPT TO RESTORE POWER TO DE-ENERGIZED BUS WOG BASIS See step 3 above. RNP DIFFERENCES/REASONS There are essentially no differences. Rev 18 Page 8 of 951 SD-024 CONTAINMENT SPRAY SYSTEM

1.0 INTRODUCTION

Containment Spray System (CSS) flow serves to reduce the containment pressure and temperature following a LOCA or main steam line break (MSLB) to near normal conditions. In addition, the CSS flow removes radioactive iodine from the containment atmosphere following a LOCA to limit off-site exposures to within the guidelines of lOCPR Part 100. The components of the CSS include the containment spray (CS) pumps, Spray Additive Tank (SAT), eductors, spray ring headers and nozzles, instrumentation and the necessary piping and valves. In addition, the CSS utilizes the Refueling Water Storage Tank (RWST) during the post-accident injection phase and the RHR System (pumps, heat exchangers and associated valves and piping) for long term post-accident recirculation phase of containment spray. 2.0 GENERAL DESCRIPTION 2.1 System Purpose The primary purpose of the CSS is to spray cool water into the containment atmosphere when appropriate in the event of a LOCA or MSLB and thereby ensure that containment pressure does not exceed its design value which is 42 psig at 263°P. This protection is afforded for all pipe break sizes up to and including the double ended rupture of the largest RCS pipe or a complete severance of a main steam line. Although the water in the core after a LOCA is quickly subcooled by the Safety Injection System, the CSS design is based on the conservative assumption that the core residual heat is released to the containment as steam. A second purpose served by the CSS is to remove elemental iodine from the containment atmosphere should it be released in the event of a LOCA. Design Basis css The design basis provides sufficient heat removal capacity the post-accident pressure below the desig!1 pressure, assuming that the core residual heat is released to the containment as steam. Adequate post-accident containment heat removal capability utilizes two separate, full-capacity, engineered safety feature (ESP) systems. One system being the CSS, and the second being the Containment Air Recirculation Cooling Ventilation System (see SD-036 HVAC Systems). These two systems serve as independent backups to each other for containment heat removal functions. Either system is fully capable of addressing post-accident containment heat loads One of the . Spray pumps is at least equivalent to two fan-coolers for heat removal capability. Page 7 of26 Revision 9 INFORMATION USE ONLY 1 . 026 A2.04 0011/1/1/// QUESTIONS REP9RT for AUDIT (:! 00 ( ) Given the following conditions:

• A reactor trip and safety injection have occurred.
• The crew is performing actions of PATH-1.
• CV Spray is manually initiated on rising Containment pressure.
• CV Spray Pump "A" breaker has a trip indication.
• CV Spray Pump "B" is operating normally with required flow.
• All other ECCS equipment is operating as required Which ONE (1) of the following is the impact on the plant and the action required?

A. Containment pressure may exceed design. Immediately transition to FRP-J.1, Response to High Containment Pressure, due to the RED condition. B:I Containment pressure will not exceed design. Attempt one reset and restart of "A" CV Spray Pump Breaker. Continue in PATH-1 until SPDS is reset and transition to another procedure is required. C. Containment pressure may exceed design. Attempt one reset and restart of "A" CV Spray Pump breaker. If restart unsuccessful, transition to FRP-J.1, Response to High Containment Pressure. D. Containment pressure will not exceed design. Do NOT attempt reset and restart of "A" CV Spray Pump unless entry to FRP-J.1 is required on transition from PATH-1. B is correct. Pressure will not exceed design if 1 train of spray is available, eliminating A and C. D is incorrect because the crew may elect to reset and restart CV Spray prior to entry to FRP-J.1 (Which will not be required anyway) SRO Question 089 Tier 2 Group 1 KIA Importance Rating -SRO 4.2 Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations: Failure of spray pump. Reference(s) -PATH-1 Proposed References to be provided to applicants during examination -None Learning Objective -Question Source -New Question History -Question Cognitive Level -Comprehension 10 CFR Part 55 Content -43.5 Comments -Tuesday. June 10, 2008 9:02:43 AM 1 HLC-08 NRC Written Exam 39. Given the following: -A SGTR has occurred and the crew is performing actions lAW PATH-2. -Prior to RCS cool down to target temperature, a NOTE warns against exceeding Steam Line flow rates in excess of the High Steam Line flow setpoint. What condition is the NOTE written to prevent? A. RCS cool down rate from being exceeded. B. An SI initiation on High Steam Line Delta P. C. An SI initiation on High Steam flow. D. MSIVs isolation. 39

HLC-08 NRC Written Exam 39. 03902.4.20 OOllMAINIREHEAT STEAM/2/1/3.S/4.3/ROILOW/NINNEW -200S/PATH-2-003 Given the following: -A SGTR has occurred and the crew is performing actions lAW PATH-2. -Prior to RCS cooldown to target temperature, a NOTE warns against exceeding Steam Line flow rates in excess of the High Steam Line flow setpoint. What condition is the NOTE written to prevent? A. RCS cooldown rate from being exceeded. B. An SI initiation on High Steam Line Delta P. C. An SI initiation on High Steam flow. MSIVs isolation. The correct answer is D. A: Incorrect -Rapid ReS cooldown is expected during this event, cooldown rates will be exceeded, but the NOTE does NOT address this concern. S: Incorrect -SI setpoint will be exceeded, however SI is blocked lAW actions of PATH-2 prior to commencing cooldown. C: Incorrect -SI setpoint will be exceeded, however SI is blocked lAW actions of PATH-2 prior to commencing cooldown. D: Correct -MSIVs will isolate if the setpoint is exceeded, potentially stopping the cooldown if the crew is using Condenser Steam Dumps. Exam Question Number: 39

Reference:

PATH-2, Grid locations D-2 and D-3; PATH-2 SD, Pages 69-70. KA Statement: Knowledge of the operational implications of EOP warnings, cautions, and notes. History: New -Written for HLC-08 NRC Exam. 45 RNP WOG BASIS/DIFFERENCES STEP STEP WOG BASIS (continued) KNOWLEDGE:

• It is not intended for the operator to reevaluate the required core exit temperature or precisely interpolate between values listed in the table.
• When the required core exit temperature is reached, the intact steam generator pressure (or feed flow to a faulted steam generator) should be controlled to maintain that temperature.
• Cooldown of the RCS should be completed before continuing in the guideline.
• Natural circulation flow in the ruptured loops may stagnate during this cooldown.

The hot leg temperature in that loop may remain significantly greater than the intact loops. In addition, safety injection flow into the cold leg may cause the cold leg fluid temperature to decrease rapidly in that same loop. Steps to depressurize the RCS and terminate SI should be performed as quickly as possible after the cooldown has been completed to minimize possible pressurized-thermal shock of the reactor vessel.

• RCS cooldown should proceed as quickly as possible and should not be limited by the 1 OO°F/hr Technical Specification limit. Integrity limits should not be exceeded since the final temperature will remain above 350°F.
• The RCP trip criteria (Step 1) does not apply after a controlled cooldown is initiated.
• If more than one steam generator is ruptured, the lowest ruptured steam generator pressure should be used to determine the required core exit temperature.

If cooldown to a target core exit temperature is already in progress when a subsequent SGTR is diagnosed the operator should stop the cool down until the subsequent ruptured steam generator is isolated since continuing the cooldown would lower the pressure in the newest ruptured steam generator and result in unnecessary releases prior to its isolation from the intact steam generators. The target core exit temperature should be reexamined to determine if the temperature should be reduced based on the subsequent ruptured steam generator pressure. If aRCS depressurization is in progress, although it does not impact the pressure in the newest ruptured steam generator, for the sake of simplicity it should be stopped and the plant stabilized by the operator until the newest ruptured steam generator is isolated. RNP DIFFERENCES/REASONS There are no significant differences. SSD DETERMINATION This is not an SSD. D2 2N6 RNP STEP NOTE: AFTER THE LOW STEAMLINE PRESS SI SIGNAL IS BLOCKED, MAIN STEAM ISOLATION WILL OCCUR IF THE HI STEAM FLOW RATE SETPOINT IS EXCEEDED I PATH-2-BD Rev. 17 Page 69 of 110 I RNP WOG BASIS/DIFFERENCES STEP STEP D2 6 WOGBASIS PURPOSE: BASIS: To alert the operator to the potential for inadvertent steam line isolation during the subsequent steam generator depressurization. An automatic protection feature is provided to close the main steamline isolation valves when the steam pressure rate signal is exceeded. In the following step, the operator is instructed to dump steam from the intact steam generators which may result in exceeding the rate setpoint. Therefore, this note is intended to alert the operator of this possibility. KNOWLEDGE: The rapid cooldown should be continued using the atmospheric steam dumps if MSIV closure occurs. RNP DIFFERENCES/REASONS The RNP note is worded in the form of a high flow instead of a pressure rate of decrease to reflect the RNP signal. There are no significant differences. SSD DETERMINATION This is not an SSD. RNPSTEP DETERMINE THE REQUIRED CORE EXIT TEMP WOG BASIS See above RNP DIFFERENCES/REASONS There are no significant differences. SSD DETERMINATION This is not an SSD. D3 1N6 RNP STEP I PATH-2-BD WHEN PZR PRESS DECREASES TO LESS THAN 2000 PSIG, THEN BLOCK PZR PRESS/HI STEAMLINE DP SIGNALS Rev. 17 Page 70 of 110 I HLC-08 NRC Written Exam 40. Given the following: -The plant is operating at 65% RTP. -The Feedwater System is in a Normal lineup. -The following Main Feedwater Pump "A" parameters are reported: -Feed pump suction pressure is 225 PSIG. -DP indication on FS-1444A equates to Main Feedwater Pump Flow of 2850 GPM. What is the expected response of the Feedwater System and what actions are required? MFP "A" will ... A. trip on low suction pressure coincident with low flow; enter AOP-01 0, MAIN FEEDWATER/CONDENSATE MALFUNCTION. B. trip on low suction pressure ONLY; enter AOP-01 O. C. remain running; verify HCV-1459, FEEDWATER HEATER BYPASS is OPEN. D. remain running; verify FCV-1444, MFP "A" RECIRC VALVE is OPEN. 40 HLC-08 NRC Written Exam 40.059 A2.05 OOlIMAINFEEDWATERl2/1/3.1/3.4IROILOW/N/A/NEW -200S/FW-007 Given the following: -The plant is operating at 65% RTP. -The Feedwater System is in a Normal lineup. -The following Main Feedwater Pump "A" parameters are reported: -Feed pump suction pressure is 225 PSIG. -DP indication on FS-1444A equates to Main Feedwater Pump Flow of 2850 GPM. What is the expected response of the Feedwater System and what actions are required? MFP "A" will ... A'1' trip on low suction pressure coincident with low flow; enter AOP-01 0, MAIN FEEDWATERICONDENSATE MALFUNCTION. B. trip on low suction pressure ONLY; enter AOP-01 O. C. remain running; verify HCV-1459, FEEDWATER HEATER BYPASS is OPEN. D. remain running; verify FCV-1444, MFP "A" RECIRC VALVE is OPEN. The correct answer is A. A: Correct -Symptoms of a FWP suction line break. Both setpoints needed for FWP trip. Enter AOP-01 O. B: Incorrect -Need coincident signals (both low pressure and low flow) to trip MFP. C: Incorrect -MFP trips, therefore first part is incorrect. HCV-1459 OPENS at 300 PSIG MFP suction pressure. HCV-1459 will be OPEN in an attempt to supply the MFP more flow. D: Incorrect -MFP trips, therefore first part is incorrect. FCV-1444 OPENS at 1475 GPM and CLOSES at 3100 GPM. Exam Question Number: 40

Reference:

APP-007-B3 and D5; AOP-010, Page 3. KA Statement: Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Rupture in the MFW suction or discharge line. History: New -Written for HLC-08 NRC exam Tuesday, June 17, 2008 1 :21 :11 PM 46 ALARM FW PMP A LO sueT PRESS TRIP AUTOMATIC ACTIONS 1. Feedwater Pump "A" trips CAUSE 1. Loss of Condensate Pump 2. Condensate System pipe break OBSERVATIONS

1. Pump Breaker Status Lights 2. Steam Generator Level Trends 3. Condensate Header Pressure ACTIONS APP-007-B3

1. IF FW PUMP "A" has tripped AND the Main Generator is in parallel with the grid, THEN refer to AOP-010. 2. IF FW PUMP "A" has tripped AND the Unit is shutdown, THEN perform the following:

1) IF required, THEN feed the S/Gs using AFW Pump(s) 2) IF the cause of the trip is NOT known, THEN dispatch personnel to inspect the pump for indications of the cause. DEVICE/SETPOINTS

1. PSL-1497-1 1235 psig* 2. FS-1444A I 3100 gpm --> *Both low pressure and flow must be present to cause pump trip. POSSIBLE PLANT EFFECTS 1. Plant Trip REFERENCES

1. AOP-010, Main Feedwater/Condensate Malfunction

2. CWD B-190628, Sheet 615, Cable M I APP-007 Rev. 35 Page 16 of 571 ALARM FW HDR LO PRESS AUTOMATIC ACTIONS 1 . None Applicable CAUSE 1. Loss of Feedwater Pump 2. Excessive load for number of FW Pumps operating

3. System break/leakage

4. Heater Drain Pump trip 5. Malfunction of LCV-1530A, HOT LEVEL CONTROL VALVE OBSERVATIONS

1. Main Feedwater Header Pressure (PI-1420)

2. Steam Generator Level trends ACTIONS 1. IF a Feedwater transient is indicated, THEN refer to AOP-01 O. APP-007-05

2. IF power level has been increased too high with one FW Pump, THEN start an additional Pump. 3. IF a system break has occurred, THEN perform the following:

1) Attempt to isolate the break. 2) IF the break can NOT be isolated, THEN perform one of the following:

a. IF conditions permit, THEN shutdown the unit and isolate the leak. b. IF the Unit can NOT be safely shutdown, THEN trip the plant and refer to EOP Network. OEVICE/SETPOINTS

1. PSL-1420 I 925 psig POSSIBLE PLANT EFFECTS 1. Plant Trip REFERENCES

1. AOP-010, Main Feedwater/Condensate Malfunction

2. CWO 8-190628, Sheet 602, Cable J 3. ESR 01-00184, FW Header Low Pressure Alarm Setpoint Change 4. EOP Network I APP-007 Rev. 35 Page 34 of 571 Rev. 25 AOP-010 MAIN FEEDWATER/cONDENSATE MALFUNCTION Page 3 of Purpose and Entry Conditions (Page 1 of 1) 1. PURPOSE This procedure provides the instructions to mitigate abnormal conditions in the Main Feedwater and Condensate System. including Main FWP trips. Condensate Pump trips. Heater Drain Pump trips. control system malfunctions and valve failures.

2. ENTRY CONDITIONS Any abnormal condition in Main Feedwater/Condensate system resulting in a system flow transient.

with the exception of Instrument Failures. -END -21 HLC-08 NRC Written Exam 41. Given the following: -A Reactor Trip has occurred and the crew has entered PATH-1. -All S/G levels indicate Off-Scale LOW. -The SDAFW Pump is running. -FCV-6416, SDAFW PUMP DISCH FLOW VALVE is CLOSED. -MDAFW Pump "A" has tripped. -MDAFW Pump "8" flow indicates 200 GPM. Which ONE (1) of the following describes the AFW system status in support of minimum Heat Sink requirements? FCV-6416 is ... A. NOT in its intended position and must be throttled OPEN to meet minimum heat sink requirements. B. NOT in its intended position, but minimum heat sink requirements are met by MDAFW Pump "B". C. in its intended position; FCV-1425, AFW PUMP B DISCH FLOW VALVE must be throttled OPEN to establish minimum heat sink requirements. D. in its intended position; minimum heat sink requirements are met by MDAFW Pump "B". 41 HLC-08 NRC Written Exam 41.061 A3.01 001lAFW/2/114.2/4.2IRO/HIGH/N/A/RNP AUDIT -2001lAFW-008 Given the following: -A Reactor Trip has occurred and the crew has entered PATH-1. -All S/G levels indicate Off-Scale LOW. -The SDAFW Pump is running. -FCV-6416, SDAFW PUMP DISCH FLOW VALVE is CLOSED. -MDAFW Pump "A" has tripped. -MDAFW Pump "B" flow indicates 200 GPM. Which ONE (1) of the following describes the AFW system status in support of minimum Heat Sink requirements? FCV-6416 is ... A'I NOT in its intended position and must be throttled OPEN to meet minimum heat sink requirements. B. NOT in its intended position, but minimum heat sink requirements are met by MDAFW Pump "B". C. in its intended position; FCV-1425, AFW PUMP B DISCH FLOW VALVE must be throttled OPEN to establish minimum heat sink requirements. D. in its intended position; minimum heat sink requirements are met by MDAFW Pump "B". The correct answer is A. A: Correct -When S/G levels decrease following a trip, the normally open SDAFW Pump discharge valve should throttle to maintain 500 GPM. It should NOT be closed, and minimum heat sink requirements for S/G levels off-scale low are at least 300 G PM total. B: Incorrect -Minimum heat sink will NOT be met until AFW flow is increased to at least 300 GPM. C: Incorrect -FCV-6416 should be OPEN maintaining 500 GPM.FCV-1425 could be throttled to meet minimum heat sink, but FCV-6416 is NOT in its intended position. D: Incorrect -FCV-6416 should be OPEN maintaining 500 GPM. Minimum heat sink is NOT met with ONLY 200 GPM flow. Tuesday, June 17, 2008 1 :21 :11 PM 47 HLC-08 NRC Written Exam Exam Question Number: 41

Reference:

Heat Sink CSFST, SO-042, AFW, Page 15. KA Statement: Ability to monitor automatic operation of the AFW, including: AFW startup and flows. History: Tuesday, June 17, 20081:21:11 PM 48 I CSFST ENTER GREEN CSF-SAT NO NO NO CSF-3, HEAT SINK YELLOW GO TO FRP-H.5 YELLOW GO TO FRP-H.4 Rev. 4 NO YES YELLOW GO TO FRP-H.3 YELLOW GO TO FRP-H.2 RED GO TO FRP-H.1 Page 5 of 91 SD-042 AFWSYSTEM 3.3 Control Valves The AFW system contains an automatic electro-hydraulic flow control valve for each MDAFW pump and the SDAFW pump. Each valve's controller is located in the control room and all other components are located in proximity to their respective pump. These valves and their associated controls are used to set AFW pump discharge flowrate and automatically maintain the rate as S/G pressure varies. These valves provide flow control for the AFW system and NPSH (anti-cavitation) protection for the pumps. These control valves have electro-hydraulic actuators which can be automatically positioned based upon the respective AFW pump discharge flow. A local manual operator is provided for operating the control valve in the event that the control system fails. The control system can also be operated in manual from the RTGB. If RCS temperature is greater than 350°F, then using manual places the plant in a Tech Specs LCO action statement (3.7.4). With normal plant conditions, manual is used to fill the Steam Generators. MDAFW pump discharge flow control valves (1424 and 1425) control the flow from each MDAFW pump to the S/Gs. These normally closed valves begin to open when the MDAFW pumps are started. The valves "fail-closed" on loss of electric power or loss of control signal. FCV-1424 is powered from IB#2 and FCV-1425 from IB#3. When the RCS temperature is > 350°F, each control valve is normally in AUTO and set at 325 gpm. When the RCS temperature is =s 350°F, these controllers shall be in Auto and set to a flow rate of 100 gpm to ensure reliable flow control as system pressure varies. '---? SDAFW pump discharge flow control valve (6416) controls the flow from the SDAFW pump to the S/Gs. This normally open valve begins to adjust when the SDAFW pump is started. This valve will "fail-open" on a loss of electrical power or loss of the control signal. FCV-6416 is powered from LP-26. In modes 1, 2 and 3, FCV-6416 is normally in AUTO and set at 500 gpm. 4.0 INSTRUMENTATION 4.1 AFW Flow Indication System AFW There are three dual flow edge meters 500 gpm, one per S/G for the MDAFW pumps and SDAFW pump, located on RTGB. S/G 1 AFW Flow -Motor Driven, FI-1425A; S/G 2 AFW Flow -Motor Driven, FI-1425B; S/G 3 AFW Flow -Motor Driven, FI-1425C; Page 15 of 38 Steam Driven, FI-1426A Steam Driven, FI-1426B Steam Driven, FI-1426C Revision 11 INFORMATION USE ONLY 061 A2.05 0011////1// QUESTIONS REPORT for AUDIT )CJo{ Given the following conditions:

• A reactor trip has occurred.
• The crew has entered PATH..;1 and is preparing to reset SPDS and transition to the appropriate procedure.
• All SG levels indicate Off-Scale Low
• The SDAFW Pump is running with discharge valve FCV-6416 closed. * "A" MDAFW Pump tripped. * "8" MDAFW Pump flow indicates 200 GPM with discharge valve FCV-1425 open. Which ONE (1) of the following describes the event in progress and the action required?

A'I FCV-6416 has failed and must be throttled open to meet minimum heat sink requirements.

8. FCV-6416 has failed and should be opened, but minimum heat sink requiremnents are met by 18" MDAFW Pump. C. FCV-6416 is in its intended position; FCV-1425 must be throttled open to establish minimum heat sink requirements.

D. FCV-6416 is in its intended position; minimum heat sink requirements are met by "8" MDAFW Pump A is correct. When SG levels decrease following a trip, the normally open SDAFW discharge valve should throttle to maintain SG levels. It should not be closed, and minimum heat sink requirement for SG levels off-scale low are at least 300 gpm total. Common Question 018 Tier 2 Group 1 KIA Importance Rating -RO 3.1 1 SRO 3.1 Ability to (a) predict the impacts of the following malfunctions or operations on the AFW: and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations: Automatic control malfunction. Reference(s) -Heat Sink CSFST, AFW SD Proposed References to be provided to applicants during examination -None Learning Objective -Question Source -New Question History -Question Cognitive Level -Comprehension 10 CFR Part 55 Content -41/43 Comments -Category 1: Category 3: Category 5: Category 7: Tuesday, June 17, 200812:56:47 PM Category 2: Category 4: Category 6: Category 8: 1 HLC-08 NRC Written Exam 42. Which ONE (1) of the following describes the power supplies for the Circulating Water Pumps? A. CW Pump "A" -4KV Bus 1 CW Pump "B" -4KV Bus 2 CW Pump "C" -4KV Bus 3 B. CW Pump "A" -4KV Bus 1 CW Pump "B" -4KV Bus 3 CW Pump "C" -4KV Bus 5 C. CW Pump "A" -4KV Bus 2 CW Pump "B" -4KV Bus 3 CW Pump "C" -4KV Bus 4 D. CW Pump "A" -4KV Bus 1 CW Pump "B" -4KV Bus 4 CW Pump "C" -4KV Bus 5 42 HLC-08 NRC Written Exam 42. 062 K2.01 OOllAC ELECTRICAL DIST/2i1/3.3/3.4fRO/LOW/NINNEW -2008/CW-007 Which ONE (1) of the following describes the power supplies for the Circulating Water Pumps? A. CW Pump "A" -4KV Bus 1 CW Pump "B" -4KV Bus 2 CW Pump "C" -4KV Bus 3 B. CW Pump "A" -4KV Bus 1 CW Pump "B" -4KV Bus 3 CW Pump "c" -4KV Bus 5 C. CW Pump "A" -4KV Bus 2 CW Pump "B" -4KV Bus 3 CW Pump "C" -4KV Bus 4 CW Pump "A" -4KV Bus 1 CW Pump "B" -4KV Bus 4 CW Pump "C" -4KV Bus 5 The correct answer is O. A: Incorrect -CW Pump "A" is powered from 4KV Bus 1, CW Pump "B" is powered from 4KV Bus 4, CW Pump "C" is powered from 4KV Bus 5. B: Incorrect -CW Pump "A" is powered from 4KV Bus 1, CW Pump "B" is powered from 4KV Bus 4, CW Pump "C" is powered from 4KV Bus 5. C: Incorrect -CW Pump "A" is powered from 4KV Bus 1 , CW Pump "B" is powered from 4KV Bus 4, CW Pump "C" is powered from 4KV Bus 5. 0: Correct-CW Pump "A" is powered from 4KV Bus 1, CW Pump "B" is powered from 4KV Bus 4, CW Pump "G" is powered from 4KV Bus 5. Exam Question Number: 42

Reference:

EOP-001, Pages 4,7 and 8; SO-057, GW, Page 10. KA Statement: Knowledge of bus power supplies to the following: Major system loads. History: New -Written for HLC-08 NRC Exam. Tuesday, June 17, 20081:21:12 PM 49 1.0 4160V AC Buss No.1 Location: 4160V Switchgear Room Power Supply: As per RTGB Line Up Loads: Cubicle Breaker CWD Reactor Coolant Pump "A" 1 52/1 109 Circulating Water Pump "A" 2 52/2 811 Feedwater Pump "A" 3 52/3 615 Station Service Transformer 2B 4 52/4 933 Heater Drain Pump "A" 5 52/5 625 Condensate Pump "A' 6 52/6 605 Unit Aux to 4KV Bus 1 7 52/7 926 PTs and Fan Equipment 8 N/A 948 PTs and Fan Equipment and Metering 9 N/A 948 4KV Bus 1 -2 Tie 10 52/10 928 I EDP-001 Rev. 4 Page 4 of 81 4.0 4160V AC BUSS NO.4 Location: 4160V Switchgear Room Power Supply: As per RTGS Line Up Loads: Cubicle Sreaker CWD 4KV Sus 3 -4 Tie 19 52/19 931 Unit Aux to 4KV Sus 4 20 52/20 930 PTs and Fan Equipment 21 N/A 949 Condensate Pump "S" 22 52/22 606 Circulating Water Pump "S" 23 52/23 813 Feed to 4KV Sus 5 24 52/24 1344 Heater Drain Pump "S" 25 52/25 626 Feedwater Pump "S" 26 52/26 620 Reactor Coolant Pump "S" 27 52/27 101 Station Service Transformer 20 28 52/28 1041 I EOP-001 Rev. 4 Page 7 of 81 5.0 4160V AC BUSS NO.5 Location: Turbine Bldg., 1 st Level, Grid Location 3B Power Supply: As per RTGB Line Up Loads: Cubicle Breaker CWD 4KV Bus 4 to 4KV Bus 5 29 N/A 1344 PTs and Control Power Transformer 30 N/A N/A SPARE 31 52/31 N/A Station Service Transformer 2E 32 52/32 1399 Circulating Water Pump "C" 33 52/33 815 SPARE 34 52134 N/A I EDP-001 Rev. 4 Page 8 of 81 SD-057 CIRCULATING WATER SYSTEM With one CWP in operation, it will pass approximately 200,000 gpm. Consequently, the amount of sodium hypochlorite flow per intake bay will increase as the number of CWPs in service decreases. When used to chlorinate the SW system, the "A" pump injects hypochlorite solution into the 30" SW headers downstream of the automatic strainers. The amount and frequency of injection is dependent upon SW system flow rates and is determined by Chemistry personnel. 3.0 COMPONENT DESCRIPTION 3.1 Circulating Water Pumps Manufacturer Type Model Total Head Flow -3 pumps Total Flow -3 pumps Motor Rating RPM Westinghouse Vertical 90MT 35.3 feet 160,700 GPM each 482,100 GPM 1750 HP 220 There are three identical vertical single-stage CWPs. For three pump operation, they are designed to provide approximately 160,000 gpm each with a 35 ft. TDH at a peak efficiency of 85 %. Based upon the CW system resistance, single pump operation can provide approximately 200,000 gpm with a 22.5 ft TDH at an efficiency of 72 %. Gland seal and bearing cooling water is supplied (15 gpm) to each CWP from the SW system and returns to Lake Robinson. A backup supply is available from the Unit No. 1 potable water. Water is drawn from Lake Robinson and passes through the traveling screens. Two concrete walls separate the intake into three bays. A single CWP is located in each bay. The intake structure is Seismic Class I and is therefore not subject to collapse under the design earthquake loading. CWP "A" is powered from 4160V Bus No.1, CWP "B" is powered from 4160V Bus No.4 and CWP "C" is powered from 4160V Bus No.5. cw Page 10 of28 Revision 4 INFORMATION USE ONLY HLC-08 NRC Written Exam 43. ONE (1) minute following a Reactor Trip, BOTH 52/8, SOUTH OCB and 52/9, NORTH OCB breakers remain CLOSED. Which ONE (1) of the following describes the failure that caused BOTH of these breakers to remain CLOSED? A. Loss of DC Bus "A". B. Loss of DC Bus "B". C. Loss of the operating Battery Charger for DC Bus "A". D. Loss of the operating Battery Charger for DC Bus "B". 43 HLC-08 NRC Written Exam 43. 063 K3.02 OOllDC ELECT DIST/21113.5/3.7IROfLOWfNINRNP AUDIT -2001lKVAC-006 ONE (1) minute following a Reactor Trip, BOTH 52/8, SOUTH OCB and 52/9, NORTH OCB breakers remain CLOSED. Which ONE (1) of the following describes the failure that caused BOTH of these breakers to remain CLOSED? A. Loss of DC Bus "A". BY' Loss of DC Bus "B". C. Loss of the operating Battery Charger for DC Bus "A". D. Loss of the operating Battery Charger for DC Bus "B". The correct answer is B. A: Incorrect -OCB 5218 and OCB 52/9 receive their control power from DC Bus B. B: Correct -OCB 52/8 and OCB 52/9 receive their control power from DC Bus B C: Incorrect -Loss of the operating Battery Charger for DC Bus B does NOT de-energize the Bus. D: Incorrect -Loss of the operating Battery Charger for DC Bus B does NOT de-energize the Bus. Exam Question Number: 43

Reference:

SD-039, KVAC, Page 11; EPP-27, Pages 3-4; EPP-27 BD, Page 4. KA Statement: Knowledge of the effect that a loss or malfunction of the DC electrical system will have on the following: Components using DC control power. History: Tuesday, June 17, 20081:21:12 PM 50 SD-039 23014 KV AC ELECTRICAL SYSTEM The two generator breakers (52/8 and 52/9) receive 125VDC control power from "B" Battery. The remairung 230KV breakers receive 125VDC control power from a battery in the 230KV Switchyard Building. OPEN: In order to receive an open indication on the RTGB all three phases of the OCB must be open. The open indication is fed from three (3) contacts that are wired in series. CLOSED: The closed indication is fed from parallel contacts from all three phases, thus if only one phase were to close the closed indication would occur. There is a protection scheme that monitors the difference between the 3 phases. If only one phase were to close the delta between the phases would be sensed and an open single to the OCB would be generated within 3 cycles. Technical Information relative to the OCBs is available at the Transmission Department Substation Maintenance Group who maintains this equipment. 3.5 Unit Auxiliary Transformer KVAC MFG TYPE RATING Primary Windings: Secondary Windings: Voltage ratio & connection Westinghouse FOA H= 44MVA, 3PHASE, 60HZ, FOA, 55'C Rise, 49.2MVA FOA, 65'C Rise 21.95/21.42120.90/20.38/19.85KV SET ON 20.38KV X,Y= 22MVA FOA 55'C Rise, 24.640MVA 65'C Rise, 4. 160KV 20.38 KV delta -2 half capacity 4.16 KV Wye The UAT is a single, three-phase step-down transformer. The input to this transformer is directly from the IPB between the main generator and the main transformer; the voltage is reduced to approximately 4160 volts. This transformer is used when the unit is on-line to carry plant auxiliary load. It could also be used with the unit shutdown provided the main generator leads were disconnected. This transformer is filled with oil to provide a medium for cooling the windings. There are two oil-circulating pumps that circulate the oil from the transformer through cooling coils and back to the transformer. Each set of cooling coils has three fans that move air through the cooling coils to remove the heat generated by the windings. When this transformer is in-service, it normally supplies power to 4160V Buses 1, 2, 4, & 5. Page 11 of39 Revision 14 INFORMATION USE ONLY Since an SI does not occur on a loss of DC Bus B, entry to the procedure will be via EPP-4, Reactor Trip. Certain actions will be completed while in EPP-4 via Foldout A. These actions are: a. Transfer of Instrument Bus 3 to MCC-8. b. Shutdown of Emergency Diesel B. STEP SPECIFIC DESCRIPTION AND RNP DIFFERENCES The following pages will provide the RNP step number and the STEP basis for each step where applicable. This is a Robinson specific EOP, therefore there is no corresponding ERG series of steps. This procedure covers an event that is not covered by the ERGs (Loss of DC). The entire procedure may be categorized as an SSD 10. The steps within this procedure will not interfere with performance of the EOPs since this procedure does not consider any other event in progress other than a loss of DC Bus "B". The loss of a DC Bus at RNP is considered "beyond design basis" and is not analyzed in the U FSAR. RNP BASIS STEP 2 3 STEP BASIS This step provides transitional direction for the subsequent step. If the Loss of DC occurred from an at power condition, the main generator output Circuit Breakers, 52/8 and 52/9, will be closed and action will be necessary to trip them. If the event started from a low power or shutdown condition the subsequent step will not be necessary. STEP BASIS On a loss of DC Bus B the North and South Generator Circuit Breakers, 52/8 and 52/9, will receive a lockout signal. Due to the loss of DC these breakers will not open. This in turn causes backup relaYing . to open other breakers to isolate the generator. In order to accomplish actions later in the procedure and to allow reclosing the backup Circuit Breakers these breakers must be opened. There are no local controls that will open the breakers without control power. There is, however, a maintenance control (for testing) at each phase of the Circuit Breakers. This feature will trip the Circuit Breakers one phase at a time. Since this function was not intended to be performed by Site Personnel the Load Dispatcher will be notified to request assistance in opening the breakers. STEP BASIS This continuous action step is provided to initiate efforts to repair the faulted DC Bus. It is placed early in the procedure so that efforts can be made to contact Maintenance personnel. The high level step provides direction to diagnose the cause and provides transitional guidance. There are three possible failure mechanisms that are the most likely causes:

• Fault on B Battery
• Fault on B Battery Bus
• Fault on MCC-6 The failure, or tripping, of the in-service Battery Charger, is not a likely cause of the loss of DC since warning would be provided via an annunciator with ample time for Operator action to transfer the Chargers.

N4 STEP BASIS The note reminds the Operator that AFW Pump B will not be available due to a loss of Control Power. 4 STEP BASIS This step assures the maintenance of the secondary heat sink by maintaining S/G level at the standard range used throughout the EOP Network. In this case AFW Pump A and the SDAFW pump are specified since AFW Pump B is lost. I EPP-27-BD Rev 10 Page 4 of 91 Rev. 10 EPP-27 LOSS OF DC BUS "B" .. _ .. -"._-_." .... -._--Page 3 Purpose and Entry Conditions (Page 1 of 1) 1. PURPOSE The purpose of this procedure is to provide instructions in the event of a loss of DC Bus B. This procedure is applicable' anytime the EPPs are applicable (greater than 350°F). 2. ENTRY CONDITIONS This procedure is entered on indication of a loss of DC Bus B from the following procedures:

a. EPP-4. Reactor Trip Response b. EPP-7. SI Termination

-END -of 29 Rev. 10 EPP-27 LOSS OF DC BUS "B" INSTRUCTIONS

1. Check Electrical Status At Time DC Power Was Lost -AUXILIARIES ON THE UNIT AUXILIARY TRANSFORMER

2. Inform Load Dispatcher Of The Following:

a. A loss of DC Control Power has occurred b. Switchyard OCBs 52/8 AND 52/9 have failed to trip c. Backup relaying has caused a North Bus Lock Out which tripped the following Switchyard Circuit Breakers:

• 52/3. ROCKINGHAM 230KV
• 52/6. #1 230-115KV BANK 230KV
• 52/7. DARLINGTON SCPSA 230KV
• 52/12. DARLINGTON COUNTY PLANT SOUTH 230KV
• 52/14. DARLINGTON COUNTY PLANT NORTH 230KV d. Request that Load Dispatcher send a Maintenance Crew to locally trip OCBs 52/8 AND 52/9
• 3. Check The Cause Of The DC Bus Failure -KNOWN Page 4 of RESPONSE NOT OBTAINED Go To Step 3. WHEN the cause is determined.

THEN notify Maintenance to correct the problem. Observe the NOTE prior to Step 4 and Go To Step 4. 29 QUESTIONS REPORT) for AUDIT ('2 6 0( 1. 062 Kl.03 001lEDPI062 K1.03/3.5 14.0/3/18/03 ROM/COMPREHENSION//! Given the following plant conditions: One minute following a reactor trip, BOTH 230KV generator output breakers remain closed Which ONE (1) of the following describes the failure that caused both of these breakers to remain closed? A. Loss of "A" DC Bus Loss of "B" DC Bus C. of the running Battery Charger for the "A" DC Bus D. Loss of the running Battery Charger for the "B" DC Bus B is correct. a. OCB 52/8 and OCB 52/9 their control power from plant DC. Comes from DC Bus "B". c. Most 230KV breakers receive their control power from a battery in the switchyard. OCB52/8 and OCB 52/9 receive their control power from DC Bus "B". d. If there is a failure of 86P, breakers would still open on 86BU Common Question 019 Tier 2 Group 1 KIA Importance Rating -RO 3.5 / SRO 3.5 Knowledge of the physical connections and/or cause-effect relationships between the AC distribution system and the following systems: DC distribution. Reference(s) -KVAC SO Proposed References to be provided to applicants during examination -None Learning Objective -Question Source -Modified Bank Question History -KVAC010-6 Question Cognitive Level-Comprehension 10 CFR Part 55 Content -41 Comments -Category 1: EDP Category 2: 062 Kl.03 Category 3: 3.5 14.0 Category 4: 3/18/03 ROM Category 5: COMPREHENSION Category 6: Category 7: Category 8: Tuesday, June 10, 2008 9:30:52 AM 1 HLC-08 NRC Written Exam 44. Given the following: -The plant is operating at 100% RTP. -480V Bus E-1 Main Breaker trips on overcurrent. -EDG "A" starts and re-energizes 480V Bus E-1. -The EDG "A" Jacket Water Coolant Pump shaft shears during EDG start. -APP-01 0-F2, EDG A COOL WTR HIILO TEMP illuminates. -APP-01 0-A2, EDG A TROUBLE illuminates. Which ONE (1) of the following is the impact of the pump shaft failure on EDG Automatic operation? EDG "A" will .... A. shutdown when coolant water pressure decreases to 12 PSIG. B. shutdown when coolant water temperature reaches 205 OF. C. continue to operate due to EDG engine trips being defeated. D. continue to operate due to run priority from the Undervoltage start. 44 HLC-08 NRC Written Exam 44. 064 G2.1.7 OOllEMERG DIESEL GEN/211/4.4/4.7IRO/HIGH/N/AINEW -200SIEDG-007 Given the following: -The plant is operating at 100% RTP. -480V Bus E-1 Main Breaker trips on overcurrent. -EDG "A" starts and re-energizes 480V Bus E-1. -The EDG "A" Jacket Water Coolant Pump shaft shears during EDG start. -APP-01 0-F2, EDG A COOL WTR HilLa TEMP illuminates. -APP-01 0-A2, EDG A TROUBLE illuminates. Which ONE (1) of the following is the impact of the pump shaft failure on EDG Automatic operation? EDG "A"wi" .... A. shutdown when coolant water pressure decreases to 12 PSIG. B. shutdown when coolant water temperature reaches 205 of. continue to operate due to EDG engine trips being defeated. D. continue to operate due to run priority from the Undervoltage start. The correct answer is C. A: Incorrect -With the EDG in a standby condition, the engine trips (12 PSIG is an actual engine trip setpoint) are defeated to prevent spurious trips from affecting the EDG operation during emergency conditions. B: Incorrect -With the EDG in a standby condition, the engine trips (205 of is an actual engine trip) are defeated to prevent spurious trips from affecting the EDG operation during emergency conditions. C: Correct -Diesel engine trips are defeated to prevent spurious trips during emergency operation and the EDG will continue to run. D: Incorrect -EDG starts on Undervoltage, however UV has no input to engine trips or run priority. Exam Question Number: 44

Reference:

OP-604, Page 8, Step 3.11 ; SD-005, EDG, Pages 30-31. KA Statement: Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. History: New -Written for HLC-08 NRC Exam. Tuesday, June 17,20081 :21 :12 PM 51 3.B 4BO VAC is available to MCC-5, MCC-6, E-1, and E-2 and the Diesel Equipment breakers are closed lAW OP-603 .. 3.9 Fire Protection System is lined up for operation lAW OP-B09. 3.10 The Supply Ventilation Fans (HVS-6 Diesel "A", HVS-5, Diesel UB U) and the Exhaust Ventilation Fans (HVE-1B, Diesel UA U / HVE-17, Diesel UB U) are operable lAW OP-906. An exception is described below. -"B" Emergency Diesel has been evaluated to remain operable with HVS-5 operable AND HVE-17 inoperable, provided the EDG "B" room door remains closed except for normal ingress and egress AND the HVE-17 discharge damper is failed open. Failing open the HVE-17 discharge damper will require compensatory actions lAW FP-012. (ESR 99-00241) 3.11 The Engine Trip Defeat Key Switch is in the TRIPS DEFEAT position. 3.12 Generator Bearing Oil level is greater than or equal to the UStopped oillevel u mark. 4.0 PRECAUTIONS AND LlMITATIONS{ TC "PRECAUTIONS AND LIMITATIONS" \f C \1 "1" } 4.1 The Emergency Diesel Generators shall meet the OPERABILITY requirements of Improved Technical Specifications LCO 3.B.1 or LCO 3.B.2. In the event that an Emergency Diesel Generator becomes inoperable, action shall be taken in accordance with Improved Technical Specifications LCO 3.B.1 or LCO 3.B.2: 4.2 The Emergency Diesel Generator fuel oil supply shall meet the OPERABILITY requirements of Improved Technical Specifications LCO 3.B.3. 4.3 The LocaVRemote Switch on the Engine Control Panel shall be left in REMOTE position at all times unless the Diesel Generator is under Clearance, Test or as directed by this procedure. When the LocaVRemote Switch position is changed, the Air Start Solenoids will momentarily deenergize. Therefore, to minimize the potential for inadvertent EDG start, the Local/Remote Switch position shall be changed using a QUICK motion. 4.4 If the engine is to be started with lube oil temperature at or below 130°F, the speed of the engine should be reduced to keep the lube oil pressure below 55 psig until the oil temperature is above 130 D F. 4.5 When the Lube Oil Filter pressure differential exceeds 10 psid, cartridge replacement is required. IOP-604 Rev. 70 Page B of 1521 SD-005 EMERGENCY DIESEL GENERATOR SYSTEM 4.2 Diesel Engine Trips 4.2.1 Engine Overspeed Mechanically trips the fuel racks at 1035-1053 rpm. When the engine is manually tripped by the emergency stop push-button, it also trips the fuel racks to the No-Fuel position. Stopping the diesel by either method dictates that the reset lever be used to reset the fuel racks. This trip cannot be defeated. 4.2.2 High Crankcase Pressure During normal operation the diesel crankcase is maintained with a slight vacuum from 0.4 to 4.0 inches of H20. If the pressure increases to +0.5 inches of H20, a trip signal is generated and stops the diesel (energizes the governor shutdown solenoid which causes the governor to reposition the fuel rack to No-Fuel position). This trip is placed in service 20 seconds after diesel starting by a time delay relay. This trip is bypassed when the Trips Defeat Switch is in the Trips Defeat position. 4.2.3 Low Lube Oil Pressure Upon diesel starting, this trip is bypassed by a time delay relay for 20 seconds to allow the diesel to come up to speed and develop enough oil pressure to reset the trip. This trip is normally set at 18 psig. If the trip setpoint is reached then the engine shuts down (energizes the governor shutdown solenoid which causes the governor to reposition the fuel rack to No-Fuel position). This trip is bypassed when the Trips Defeat Switch is in the Trips Defeat position. 4.2.4 Low Jacket Water Pressure This trip is defeated for 20 seconds after the diesel reaches 810 rpm by a time delay relay. This allows the diesel to come up to speed and stabilize. If coolant pressure decreases to approximately 12 psig during operation, a shutdown of the diesel will result (energizes the governor shutdown solenoid which causes the governor to reposition the fuel rack to No-Fuel position). This trip is bypassed when the Trips Defeat Switch is in the Trips Defeat position. "-7 4.2.5 High Jacket Water Temperature EDG During normal operation (2500 KW) cooling water temperature out of the engine should be 182°P to 185°P (alarms at 195°F Hi and 105°P Lo). If the temperature increases to 205°P, a temperature switch will shutdown the diesel (energizes the Page 30 of 74 Revision 11 INFORMATION USE ONLY SD-005 EMERGENCY DIESEL GENERATOR SYSTEM governor shutdown solenoid which causes the governor to reposition the fuel rack to No-Fuel position). This trip is placed in service 20 seconds after diesel starting by a time delay relay. This trip is bypassed when the Trips Defeat Switch is in the Trips Defeat position. 4.2.6 Engine Start Failure If the engine speed has not increased to > 200 rpm (LSR) within 10 seconds of the start signal (TD2), then a governor shutdown occurs (energizes the governor shutdown solenoid which causes the governor to reposition the fuel rack to No-Fuel position). This trip is bypassed when the Trips Defeat Switch is in the Trips Defeat position. Governor Shutdown is reset by depressing the ALARM RESET push button on the Engine Control Panel. NOTE: Diesel trips (High Crankcase Pressure, Low Lube Oil Pressure, Low Jacket Water Pressure, and ,High Jacket Water and Start Failure) are normally defeated by a key-operated switch (Trips Defeat Switch) on the diesel control panel. For normal testing the trips are placed in after the diesel has started and assumed load as allowed by Technical Specifications. This action reduces exposure of the EDGs to undue risk of damage that might render it inoperable. 4.2.7 10 Second Overcrank If the engine speed has nO,t increased to > 200 rpm (LSR) within 10 seconds of the start signal (TD2), then the air start solenoids open, preventing further loss of starting air until the malfunction causing the overcrank is corrected. There is enough stored air for eight cold starts. However, the EDG will only attempt to auto start once. This is due to the engine control system which is designed to stop cranking within 10 seconds. Failure of the engine to start within the timing period of the overcrank time (10 seconds) indicates a malfunction. Shutdown conserves the starting air supply so the engine can be subsequently started after the malfunction is corrected. This trip will not prevent a manual start of the EDG, and is not defeatable. 4.3 Generator Trips 4.3.1 Reverse Power This trip opens the generator output breaker to prevent motorizing the generator and possibly damaging the diesel. NOTE: After a generator trip from Reverse Power, when the generator output breaker trips, the trip relay is automatically reset. The flag on the associated relay on the Generator EDG Page 31 of 74 Revision 11 INFORMATION USE ONLY HLC-08 NRC Written Exam 45. Given the following: -The plant is operating at 100% RTP. -A release of Waste Condensate Tank (WCT) "A" is in progress. -APP-036-E7, RAD MONITOR TROUBLE, is received. -The BOP Operator reports the FAIL light for R-18, LIQUID WASTE DISPOSAL EFFLUENT monitor, is ON. Which ONE (1) of the following describes the status of RCV-018, LIQUID WASTE RELEASE ISOLATION VALVE? RCV-018 will ... A. NOT automatically close. The release must be terminated manually. B. automatically close when the monitor FAIL light is illuminated. C. NOT automatically close, and CANNOT be closed from the Waste Disposal Panel. D. automatically close, and must be reset by cycling the valve controller's potentiometer. 45 HLC-08 NRC Written Exam 45.073 A4.01 OOlIPROCESS RAD MONITORl2/113.9/3.9IROIIllGHlNIA/RNP AUDIT -2006IRMS-009 Given the following: -The plant is operating at 100% RTP. -A release of Waste Condensate Tank (WCT) "A" is in progress. -APP-036-E7, RAD MONITOR TROUBLE, is received. -The BOP Operator reports the FAIL light for R-18, LIQUID WASTE DISPOSAL EFFLUENT monitor, is ON. Which ONE (1) of the following describes the status of RCV-018, LIQUID WASTE RELEASE ISOLATION VALVE? RCV-018 will ... Art NOT automatically close. The release must be terminated manually. B. automatically close when the monitor FAIL light is illuminated. C. NOT automatically close, and CANNOT be closed from the Waste Disposal Panel. D. automatically close, and must be reset by cycling the valve controller's potentiometer. The correct answer is A. A: Correct -Fail light means loss of power and/or loss of indication. Valve will NOT close. Release must be terminated. B: Incorrect -RCV-018 will close on a High Radiation Alarm, NOT a fail light. C: Incorrect -RCV-018 can be operated at any time with control switch on the Waste Disposal Panel. D: Incorrect -RCV-018 will NOT Automatically CLOSE and is controlled by the CLOSE-AUTO-OPEN switch instead of potentiometer. RCV-014 is controlled by a potentiometer. Exam Question Number: 45

Reference:

SD-019, RMS, Page 40; APP-036-E7. KA Statement: Ability to manually operate and/or monitor in the control room: Effluent release. History: Tuesday, June 17,20081:21:12 PM 52 SD-019 RADIATION MONITORING SYSTEM RMS 2. Rate-meters (located in RMS consoles 1, 2, and 3) a. Nuclear Re rch Co oration, R-l throu h R-9, R-ll, R-12, R-15, R-16, R-17 R-18:' R-20 R-21 R-30, R-31A, R-31B R-31C, and R-33

• Red light -ALARM/RESET

-==,.

• Amber light -FAIL-No automatic actions occur.
• White light -PWR ON (power on)
• White light -CKT TEST (circuit test)
• White light -CHECK SOURCE
• White light -H. V. Off (High Voltage Off)
• Digital Indicator 10-1 to 10 4 mr/hr (R-l through R-S)
• Digital Indicator 10° to 10 5 mr/hr (R-9, R-30, R-31A, R-31B, R-31C and R33)
• Digital Indicator 10 1 to 10 6 cpm (R-ll, R-15, R-16, R-17, and IS)
• Digital Indicator 10 1 to 10 7 cpm (R-12, R-20, and R-21) b. Victoreen, R-32A and R-32B (located in RMS console 1)
• Yellow light -ALERT
• Red light -HIGH
• Blue light -CHANNEL TEST
• Green light -SAFE/RESET
• Red button -E.C.S. (Electronic Check Source)
• Analog Indicator 10° to 10 7 R/hr 3. Indicators, Digital (located in RMS consoles 2 and 3; FIGURES 10 & 22) a. R-14C, R-14D, and R-14E 10 1 to 10 6 cpm b. R-19A, R-19B, R-19C 10 1 to 10 7 cpm 4. Indicators, Light (located in RMS consoles 2 and 3) a. R-14C, R-14D, R-14E, R-19A, R-19B, R-19C, and R-14 Skid
• Red light -ALARM (except R-14 Skid)
• Amber light -FAIL 5. Pump Controllers, Indicator Lights (located in RMS consoles 2 and 3) a. R-ll/R-12 Page 40 of 59 INFORMATION ONLY Revision 7 ALARM RAD MONITOR TROUBLE *** WILL REFLASH *** AUTOMATIC ACTIONS APP-036-E7 Page 1 of 3 1. IF the R-14C FAIL alarm is ILLUMINATED, THEN R-14C closes RCV-014, WASTE GAS DECAY SYSTEM ISOLATION VALVE, to stop any gas release in progress.

CAUSE For all channels (R-1 through R-9; R-11; R-12; R-15; R-16; R-17; R-18; R-20; R-21; R-30; R-31 A, B, C; R-32A, B; R-33): . -Loss of Counts Loss of Power 2. For Channel R-14C, D, E: Loss of Counts Loss of Power Low Sample Flow Low F-14 Flow Low Battery 3. For PLANT VENT EFFLUENT MONITORING EQUIPMENT FAIL: F-14 Kurz Power Failure Heat Trace Trouble Stack Flow Trouble 4. For R-19A, B, C: Loss of Counts Loss of Power Low Skid Flow High Temperature Check Source Counts NOT within limits OBSERVATIONS

1. FAIL light for associated RMS channel illuminated.

2. Plant Vent Effluent Monitoring Equipment FAIL light for R-14C/D/E illuminated.

I APP-036 Rev. 62 Page 45 of 961 APP-036-E7 Page 2 of 3 ACTIONS NOTE: R-14C/D/E do NOT have to be declared out of service if only the Plant Vent Effluent Monitoring Equipment FAIL light is illuminated unless further investigation by E&C determines the channel(s) are inoperable.

1. IF the Plant Vent Effluent Monitoring Equipment FAIL light for R-14C/D/E illuminates, THEN NOTIFY E&C to investigate cause of alarm. 2. IF a channel FAIL light has illuminated, THEN PERFORM the following:

1) Attempt to RESET the alarm. 2) IF the FAIL light extinguishes, THEN channel is operable AND no further actions are required.

3) IF the FAIL light will NOT extinguish, THEN DECLARE the channel inoperable until the results of the subsequent E&C status check is known. 3. IF any channel has failed, THEN REVIEW TECH SPECS AND ODCM to determine the appropriate actions for any release in progress through affected channels with an illuminated FAIL light. 4.

6. -ODCM Table 2.6-1 for liquid releases -ODCM Table 3.10-1 for gaseous releases DECLARE any channel with an illuminated FAIL light inoperable until the cause for the FAIL light is determined.

It may be necessary to keep the affected channel energized if a continued release is allowed lAW applicable TECH SPECS AND ODCM. IF R-18 FAIL light is ILLUMINATED, THEN SECURE any release in progress via this pathway. IF R-11 OR R-12 FAIL lights are illuminated, AND a Containment Purge is in progress, THEN PERFORM the following:

1) IF the plant is in Modes 1 through 4, THEN SECURE any Containment Purge in progress.

Containment Purge is NOT allowed in Modes 1 through 4 unless R-11 AND R-12 are in service (ITS LCO 3.3;6)(ACR 94-00833).

2) IF movement of recently irradiated fuel is in progress, THEN STOP movement of that fuel. 7. IF R-1 is inoperable, AND the plant is in Modes 1 through 4 OR movement of fuel assemblies is in progress, THEN VERIFY Control Room Ventilation System is in the Pressurization Mode. (ITS LCO 3.3.7) 8. NOTIFY E&C of monitor(s) status. 9. Releases may be continued lAW applicable TECH SPECS AND ODCM. 10. IF desired, THEN REMOVE the affected channel from service using OWP-014. I APP-036 Rev. 62 Page 46 of 961 OEVICE/SETPOINTS

1. Refer to OMM-014 POSSIBLE PLANT EFFECTS 1. Entry to TECH SPEC LCO REFERENCES

1. ITS LCO 3.3.6, 3.4.15, ODCM 2.6 and 3.10 2. ACR 94-00833, No CV Purge with R-11 and R-12 OOS 3. ACR 94-01308, R-18 FAIL Condition APP-036-E7 Page 3 of 3 4. CWO B-190628, Sheets 82-85,87,279,350,361,525,530,535,637, 1058,1693-1695,1724, 1727,1728,1734,1735,1740, 1741,1741A

5. OMM-014, Radiation Monitoring Setpoints

6. ESR 95-00227 7. OWP-014, Radiation Monitoring System (RMS) 8. CR 97523, R-14C Failure Due To Low Counts 9. EC 52464, Replace R-14 Plant Vent Monitor I APP-036 Rev. 62 Page 47 of 961 QUESTIONS RElj-pRT ,.\ for AUDIT (06 ;f(jdrrJ 1. 073 A2.02 001lSD-019/073 Al.01l3.2/3.1I07/01l03MCM/MEMORY/OI-01 NRC EXAM/I Given the following plant conditions:

• The plant is in MODE 1, 100%RTP
• A release of Waste Condensate Tank (WCT) 'A' is in progress
• Annunciator APP-036-E7, RAD MONITOR TROUBLE, is received
• The BOP Operator reports the FAIL light for R-18, Liquid Waste Disposal Monitor, is ON Which one of the following describes the status of the Liquid Waste Release Isolation, RCV-018? A'I RCV-018 will not automatically close. The release must be stopped manually B. RCV-018 will automatically close when the monitor FAIL light is illuminated C. RCV-018 is not immediately affected and will close if high radiation is sensed by R-18 D. Automatic operation of RCV-018 is defeated, but the release may continue unless an actual High Radiation condition exists A. Correct.

Reference:

SD-019, APP-036-E7. Fail light means loss of power and/or loss of indication. Valve will not close. Release must be terminated. Common Question 022 Tier 2 Group 1 KIA Importance Rating -RO 2.7/ SRO 2.7 Ability to (a) predict the impacts of the following malfunctions or operations on the PRM . system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Detector failure. Reference(s) -Proposed References to be provided to applicants during examination -Learning Objective -Question Source -Question History -Question Cognitive Level -10 CFR Part 55 Content -Comments -Category 1: SD-019 Category 2: 073 Al.Ol Category 3: 3.2/3.1 Category 4: 07/01/03MCM Category 5: MEMORY Category 6: 01-01 NRC EXAM Category 7: Category 8: Tuesday, June 10,200810:07:17 AM 1 HLC-08 NRC Written Exam 46. Given the following: -The plant is operating at 100% RTP. -APP-008-E8, N SW HDR STRAINER PIT HI LVL has illuminated. -APP-008-F8, NORTH SW HDR LO PRESS has illuminated. Which ONE (1) of the following are Required Actions? A. Close V6-12D, NORTH SW HEADER ISOLATION VALVE then Go to Section F, SERVICE WATER PITS FLOODING IN INTAKE AREA. B. Close V6-12B AND V6-12C, SW X-CONN Valves, then Go to Section F, SERVICE WATER PITS FLOODING IN INTAKE AREA C. Close V6-12B AND V6-12C, SW X-CONN Valves, then Go to EPP-28, LOSS OF ULTIMATE HEAT SINK. D. Close V6-12D, NORTH SW HEADER ISOLATION VALVE AND Verify SW Pumps "C" and "0" are stopped. 46 HLC-08 NRC Written Exam 46.076 A2.01 WATERf2/l/3.S/3.7IRO/LOW/N/NNEW -2008/AOP-022-003 Given the following: -The plant is operating at 100% RTP. -APP-008-E8, N SW HDR STRAINER PIT HI LVL has illuminated. -APP-008-F8, NORTH SW HDR LO PRESS has illuminated. Which ONE (1) of the following are Required Actions? A. Close V6-12D, NORTH SW HEADER ISOLATION VALVE then Go to Section F, SERVICE WATER PITS FLOODING IN INTAKE AREA. B:' Close V6-12B AND V6-12C, SW X-CONN Valves, then Go to Section F, SERVICE WATER PITS FLOODING IN INTAKE AREA C. Close V6-12B AND V6-12C, SW X-CONN Valves, then Go to EPP-28, LOSS OF ULTIMATE HEAT SINK. D. Close V6-12D, NORTH SW HEADER ISOLATION VALVE AND Verify SW Pumps "c" and "D" are stopped. The correct answer is B. A: Incorrect -V6-12D is closed later in the procedure after closing V6-12B and V6-12C for leak isolation. B: Correct -Step 1 of AOP is an Immediate Action step. (1) Check APP-008-E8, N SW HDR STRAINER PIT HI LVL extinguished. (RNO) Then immediately close V6-12B and V6-12C. C: Incorrect -Seperation of headers by closing X-connects is correct, but only Go to EPP-28 if NO SW available. D: Incorrect -V6-12D closed and verifying pumps stopped are subsequent steps AFTER immediate actions. Exam Question Number: 46

Reference:

AOP-022, Pages 3, 4 and 38; APP-008-E8; APP-008-F8. KA Statement: Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations: Loss of SWS. History: New -Written for HLC-08 NRC Exam. Tuesday, June 17, 20081 :21 :13 PM 53 Rev. AOP-022 LOSS OF SERVICE WATER Page Purpose and Entry Conditions (Page 1 of 1) 1. PURPOSE This procedure provides instructions in the event of a break of either the North or South Service Water Headers upstream OR downstream of check valves SW-541 or SW-545 or flooding in the Intake Area Service Water Pits. 2. ENTRY CONDITIONS This procedure is entered whenever there is an indication that a break of a Service Water Header has occurred. -END -31 3 of 78 Rev. 30 AOP-022 LOSS OF SERVICE WATER Page 4 of INSTRUCTIONS RESPONSE NOT OBTAINED Step 1 is an immediate action step. 1. Check The Following Alarms -EXTINGUISHED:

• APP-008-E7.

S SW HDR STRAINER PIT HI LEVEL APP-008-E8. N SW HDR STRAINER PIT HI LEVEL

• 2. Check SW -ANY AVAILABLE

3. Make PA Announcement For Procedure Entry Perform the following: Close the following SW X-CONN Valves:

• V6-12B
• V6-12C b. Go To Section F. IF a total loss of Service Water occurs due to hostile action. THEN Go To EPP-28. Loss of Ultimate Heat Sink. A SW Header leak may be identified by observing the sequence in which SW Header low pressure alarms are received.

and evaluating SW Header pressure indications.

4. Check Leak Location -IDENTIFIED Perform local inspections as necessary to determine leak location.

WHEN the leak location is identified. THEN observe the NOTE prior to Step 5 and Go To Step 5. 78 Rev. AOP-022 LOSS OF SERVICE WATER Page INSTRUCTIONS RESPONSE NOT OBTAINED SECTION F SERVICE WATER PITS FLOODING IN INTAKE AREA (Page 1 of 10) 1. Verify PA Announcement For Procedure Entry Performed 2. Verify SW X-CONN Valves -CLOSED:

• V6-12B
• V6-12C 30 38 The source of flooding in the Intake Structure may be Service Water. Fire Water. or Intake Structure leakage. 3. Evaluate Control Room Indications AND Perform Local Inspections To Determine Source Of Flooding Prior To Continuing

4. Check Source Of Flooding -SERVICE WATER 5. Check Service Water Leak Location -ON SOUTH HEADER 6. Verify The Following:

a. SW PUMP C -RUNNING b. SW PUMP D -RUNNING c. SW PUMP A -STOPPED d. SW PUMP B -STOPPED Go To Step 26. Go To Step 14. of 78 ALARM N SW HDR STRAINER PIT HI LEVEL AUTOMATIC ACTIONS 1 . None Applicable CAUSE 1. Failure of sump pump in north service water strainer pit 2. System leakage in excess of sump pump capacity OBSERVATIONS

1. Check water level in pit. ACTIONS 1. Refer to AOP-022. DEVICE/SETPOINTS

1. LS-1652A I 1 foot above floor POSSIBLE PLANT EFFECTS APP-008-E8

1. Continued flooding could jeopardize operability of valves V6-12A, V6-12B, V6-12C, & V6-12D 2. Potential to enter TECH SPEC LCO condition REFERENCES

1. ITS LCO 3.7.7 2. AOP-022, Loss of Service Water 3. HBR2-11098, SH.11 4. CWD B-190628, Sh. 833 I APP-008 Rev. 37 Page 43 of 51 I ALARM NORTH SW HDR LO PRESS AUTOMATIC ACTIONS 1. None Applicable CAUSE 1. Loss of SW Pump(s) 2. CCW Heat exchanger Outlet Valves open too far 3. Rupture of Service Water Piping 4. Season increase in SW temperature (slow transient)

OBSERVATIONS

1. Service Water Pressure (PI-1616, PI-1684) 2. Service Water Pump Breaker(s)

Indicating Lights ACTIONS APP-008-FS

1. IF an operating SW Pump has tripped, THEN perform the following: 1 ) START a Standby Pump. 2) Dispatch operator to check breaker(s)

SW Pump A -480V Bus E1 (CMP 20B) SW Pump B -480V Bus E1 (CMP 19C) SW Pump C -4S0V Bus E2 (CMP 24A) SW Pump 0 -480V Bus E2 (CMP 25B) 3) Throttle CCW Heat Exchanger Return Valves, as necessary, to maintain 40 to 50 psig in the SW Headers. 2. IF a rupture in a SW Header has occurred, THEN refer to AOP-022. 3. IF an increase in SW temperature has caused SW cooling valves to throttle open, THEN locally throttle SW-739 AND SW-740 as necessary to maintain SW pressure 40 psig to 50 psig. DEVICE/SETPOINTS

1. PSL-1616 140 psig POSSIBLE PLANT EFFECTS 1. Loss of Service Water 2. Overheat of CCW 3. Possible entry into TECH SPEC LCO REFERENCES

1. ITS LCO 3.7.7 2. AOP-022, Loss of Service Water 3. CWO B-190628, Sheet 841, cable L 4. Flow Diagram G-190199 I APP-008 Rev. 36 Page 51 of 51 I HLC-08 NRC Written Exam 47. Given the following: -A line break on the Instrument Air header inside Containment has resulted in a full depressurization of Containment Instrument Air. -The crew has isolated Instrument Air to Containment.

Which ONE (1) of the following can be operated remotely? A. Pressurizer Spray valves. B. Pressurizer PORVs. C. Letdown Line Isolation valves. D. Charging Line Isolation valves 47 HLC-08 NRC Written Exam 47.078 Kl.03 OOllINSTRUMENT AIR/2/1/3.3/3.4IROILOW/NINNEW -2008/PZR-OlO Given the following: -A line break on the Instrument Air header inside Containment has resulted in a full depressurization of Containment Instrument Air. -The crew has isolated Instrument Air to Containment. Which ONE (1) of the following can be operated remotely? A. Pressurizer Spray valves. Pressurizer PORVs. C. Letdown Line Isolation valves. D. Charging Line Isolation valves The correct answer is B. A: Incorrect -PZR Spray Valves use Instrument Air as motive force. Valves have failed CLOSED. B: Correct -PZR PORVs use Nitrogen from SI Accumulator fill line and have a dedicated Nitrogen Accumulator for each valve. Instrument Air serves as backup. C: Incorrect -' Letdown line isolation valves use Instrument Air and ALL fail CLOSED. 0: Incorrect -CVC-310A and 310B fail OPEN. CVC-311 fails CLOSED. ALL charging line isolations are non-functional with Instrument Air isolated. Exam Question Number: 47

Reference:

AOP-017, Attachment 1, Pages 34,36; SD-059, PZR, Page 10 and Figure 4. KA Statement: Knowledge of the physical connections and/or cause-effect relationships between the lAS and the following systems: Containment air. History: New -Written for HLC-08 NRC Exam. Tuesday, June 17, 20081:21:13 PM 54 Rev. AOP-017 LOSS OF INSTRUMENT AIR Page ATTACHMENT 1 MAJOR COMPONENTS AFFECTED BY LOSS OF IA (Page 1 of 5) 1. Chemical and Volume Control System Components FAIL POSITION a. APP-003-F3. CHG PMP LO SPEED -ILLUMINATED (Loss of air to pressure switch ,downstream of IIp Convertor)

b. CHARGING PUMP SPEED CONTROL -NO FAILURE (Back-Up Air) CVC-200 A. B & C. LTDN ORIFICES -CLOSED O(SI'R. i'C' I CVC-204 A & B. LTDN LINE ISOs -CLOSED e. CVC-303 A. B & C. SEAL LEAKOFFS -OPEN f. CVC-307. PRI SEAL BYP ISO -CLOSED g. CVC-310A.

LOOP 1 HOT LEG CHG -OPEN D/o DI51R:' '-7' h. CVC -31 OB. LOOP 2 COLD LEG CHG -OPEN ....,. i. CVC -311. AUX PZR SPRAY -CLOSED j. CVC-387. EXCESS LTDN STOP -CLOSED k. CVC-389. EXCESS LTDN DIV -FAILS TO VCT 1. CVCS HUT LEVEL CONTROLLERS -FORCED LOW m. FCV-113A. BA TO BLENDER -OPEN n. FCV-113B.' BLENDED MU TO CHG SUCTION -CLOSED o. FCV-114A. PW TO BLENDER -CLOSED p. FCV -114B. BLENDED MU TO VCT -CLOSED q. HCV-I0S. BORIC ACID TK B RECIRC -CLOSED r. HCV-110. BORIC ACID TK A RECIRC -CLOSED fJ/fff"V'" s . HCV-121. CHARGING FLOW -OPEN t. HCV-137. EXCESS LTDN FLOW -CLOSED u. LCV-llSA. VCT/HLDP TK DIV -FAILS TO VCT (CONTINUED NEXT PAGE) 35 34 of 61 AOP-017 LOSS OF INSTRUMENT AIR ATTACHMENT 1 MAJOR COMPONENTS AFFECTED BY LOSS OF IA (Page 3 of 5) 6. Isolation Valve Seal Water System Components FAIL POSITION a. PCV-1922 A & B. IVSW AUTO HEADER ISOLs -OPEN 7. Main Steam System Components FAIL POSITION a. MAIN STEAM ISOLATION VALVES -CLOSED b. STEAM LINE PORVs -CLOSED 8. Primary Sample System Components FAIL POSITION a. PS-956 A through H. PRIMARY SAMPLE ISOLATIONS -CLOSED 9. Radiation Monitoring System Components FAIL POSITION a. RMS-1.2.3 & 4. R-11/R-12 ISOL VALVES -CLOSED 10. Reactor Coolant System Components FAIL POSITION a. PCV-455 A & B. PZR SPRAYS -CLOSED

b. RC-516 & 553. PRT TO GAS ANALYZER -CLOSED c. RC-519 A & B. PW TO CV ISOs -CLOSED d. RC-544. RV FLANGE LEAKOFF -OPEN e. RC-550. PRT NITROGEN SUPPLY -CLOSED 11. Residual Heat Removal System Components FAIL POSITION a. HCV-142. PURIFICATION FLOW -CLOSED b. HCV-758. RHR HX DISCH FLOW -CLOSED c. FCV-605. RHR HX BYPASS FLOW -CLOSED Rev. 35 Page 36 of 61 SD-059 PRESSURIZER SYSTEM level, which reduces the elevation head loss that spray flow must overcome.

Normal spray flow is unlikely or will not occur at all when 'C' RCP is stopped and PZR level is less than 30%. Therefore, PZR pressure response may not be as expected for the above condition. (Ref. SCR 90-031) 3.4 PZR Surge Line Nozzle Diameter Pipe Schedule Surge Line Surge Line Nominal Thickness Design Pressure Design Temperature 14 in. 140 12 in. 1.125 in. 2485 psig 680 0 P The PZR surge line, which connects the bottom of the PZR to RCS loop C hot leg, is sized such that it will pass the maximum anticipated surge flow of 20,000 gpm with a minimal pressure drop. A resistance temperature detector is installed in the surge line and provides indication and a low temperature alarm in the Control Room. Low temperature is indicative of stagnation of the PZR fluid. The surge nozzle, located in the bottom of the vessel, is protected against thermal shock by a thermal sleeve. A retaining screen above the nozzle prevents foreign matter in the PZR from entering the RCS piping. Incoming surge flow displaces the water in the vessel as it enters the heater bundle area. 3.5 PZR Safety and Relief Valves Three spring loaded safety valves and two PORVs provide for over pressure protection. The motive force for the PORVs is nitrogen with an IA backup. 3.5.1 Safety Valves (RC-551A, B, & C) (PZR-Pigure 3 and 5) Number Capacity Set Pressure Back Pressure Normal Relieving 3 293,330 lb/hr each at 3 % accumulation 2485 psig 3 psig 350 psig (maximum) The safety valves, set for the system design pressure of 2485 psig, are spring loaded, enclosed pop type, with backpressure compensation. The combined capacity of the valves is equal to, or greater than, the maximum surge rate resulting from complete loss PZR Page 10 of 27 Revision 9 INFORMATION USE ONLY PZR PORV PNEUMATIC SYSTEM PZR-FIGURE-4 INFORMATION USE ONLY TO PCYf pzrf04 av-! av-l HLC-08 NRC Written Exam 48. Given the following: -The plant is in MODE 4, cooling down lAW GP-007, PLANT COOLDOWN FROM HOT SHUTDOWN TO COLD SHUTDOWN. -A spurious R-11 alarm causes a CV Ventilation Isolation signal. -V12-8 and V12-9, CV PURGE OUTLET VALVES, both have dual OPEN/CLOSED indications. Which ONE (1) of the following, if any, is required? A. Close and deactivate EITHER V12-8 OR V12-9 within 1 hour. B. Close and deactivate BOTH V12-8 AND V12-9 within 1 hour. C. Using CV Purge Valve blocking mechanism, override EITHER V12-8 OR V12-9 CLOSED within 1 hour. D. Plant is in MODE 4; No action required. 48 HLC-08 NRC Written Exam 48. 103 K3.01 001lCONTAINMENT/2/1/3.3/3.7IRO/LOW/N/NNEW -2008/CV-008 Given the following: -The plant is in MODE 4, cooling down lAW GP-007, PLANT COOLDOWN FROM HOT SHUTDOWN TO COLD SHUTDOWN. . -A spurious R-11 alarm causes a CV Ventilation Isolation signal. -V12-8 and V12-9, CV PURGE OUTLET VALVES, both have dual OPEN/CLOSED indications. Which ONE (1) of the following, if any,.is required? At Close and deactivate EITHER V12-8 OR V12-9 within 1 hour. B. Close and deactivate BOTH V12-8 AND V12-9 within 1 hour. C. Using CV Purge Valve blocking mechanism, override EITHER V12-8 OR V12-9 CLOSED within 1 hour. D. Plant is in MODE 4; No action required. The correct answer is A. A: Correct -ITS 3.6.3, CONTAINMENT ISOLATION VALVES, Condition B applies. Either valve must be closed and deactivated within 1 hour to meet ITS requirements. B: Incorrect -Only 1 valve must be closed/deactivated in each line. C: Incorrect -Blocking mechanisms only block the valves OPEN. D: Incorrect -CV isolation required in MODES 1,2,3 and 4 for containment integrity to be met. Exam Question Number: 48

Reference:

ITS 3.6.3; OP-921, Page 49. KA Statement: Knowledge of the effect that a loss or malfunction of the containment system will have on the following: Loss of containment integrity under shutdown conditions. History: New -Written for HLC-08 NRC Exam. 55 3.6 CONTAINMENT SYSTEMS 3.6.3 Containment Isolation Valves Containment Isolation Valves 3.6.3 LCO 3.6.3 Each containment isolation valve shall be OPERABLE. APPLICABILITY: MODES 1. 2. 3. and 4. ACTIONS ------------------------------------- NOTES -------------------. ---------------- I. Penetration flow path(s) 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 containment isolation valves. 4. Enter applicable Conditions and Required Actions of LCO 3.6.1. "Containment," when isolation valve leakage results in exceeding the overall containment leakage rate acceptance criteria.

5. Enter applicable Conditions and Required Actions of LCO 3.6.8, "Isolation Valve Seal Water (lVSW) System," when required IVSW supply to a penetration flowpath is isolated . ...................

__ ......... _--_ .... --.............

---

.. -.. _ .............. . CONDITION REQUIRED ACTION COMPLETION TIME A. --------.

NOTE ---------A.l Isolate the affected 4 hours Only applicable to flow path penetration flow paths y use of at least with two containment one closed and isolation valves. de-activated ........ _ ............. automatic valve, closed manual valve, One or more blind flange, or penetration flow paths check valve with flow with one containment through the valve isolation valve secured. inoperable. AND (continued) HBRSEP Unit No. 2 3.6-7 Amendment No. 176 ACTIONS CONDITION A. (continued) A.2 -B. *********NOTE********* B.1 Only applicable to penetration flow paths with two containment isolation valves. _.----.-....... _ ...... One or more penetration flow paths with two containment isolation valves inoperable. HBRSEP Unit No. 2 Containment Isolation Valves 3.6.3 REQUIRED ACTION COMPLETION TIME ........ NOTE* ........ Isolation devices in high radiation areas may be verified by use of administrative means . ........ -........ ----Verify the affected Once per 31 days penetration flow path for isolation is isolated. devices outside containment AND Prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days for isolation devices inside containment Isolate the affected 1 hour genetration flow path y use of at least one closed"and de*activated automatic valve. closed manual valve. or blind flange. (continued) 3.6*8 Amendment No. 176 CONTINUOUS USE Section 8.4.5 Page 1 of 3 8.4.5 Supply and Exhaust Valves{ TC "Blocking Open 'p'urge Sup an Exhaust Valves" \f C \1 "3" } (CR 95-01213) NOTE: NOTE: IOP-921 1. Initial Conditions

a. This revision has been verified to be the latest revision available.

_______ (Print) _______ _ Name Signature Date 2. Instructions Refer to Attachment 10.1, Purge Valve Blocking Mechanism, for identification of component parts. Steps without sign offs may be repeated as necessary to accommodate blocking or disengaging the block of any of the purge valves. a. IF containment purge valve, THEN PERFORM the following:

1) VERIFY that the purge valve to be blocked open is open. 2) REMOVE the hex nut locking pin. 3) RAISE the hex nut sufficiently to allow the jackscrew to be inserted to engage with the valve actuator.

4) ROTATE jackscrew clockwise until it engages with the valve actuator.

5) ALIGN the hex nut with locking pin holes in the jackscrew and install locking pin. Rev. 47 Page 49 of 571 HLC-08 NRC Written Exam 49. Which ONE (1) of the following is the principal concern when APP-001-A2, SEAL WTR INJ FILTER HI DELTA-P is in alarm? A. RCPs may NOT be receiving adequate seal injection flow. B. Accumulation of contaminants on the Seal Injection Filter could cause ALARA concerns.

C. Seal Injection flow rates may be in excess of 10 GPM to each RCP. D. Total Seal Injection flow rate may exceed 20 GPM. 49 HLC-08 NRC Written Exam 49. 004 K3.08 001lCVCS/2/113.6/3.8fRO/LOWINIAfNEW -2008/CVCS-008 Which ONE (1) of the following is the principal concern when APP-001-A2, SEAL WTR INJ FILTER HI OELTA-P is in alarm? A'I RCPs may NOT be receiving adequate seal injection flow. B. Accumulation of contaminants on the Seal Injection Filter could cause ALARA concerns.

c. Seal Injection flow rates may be in excess of 10 GPM to each RCP. D. Total Seal Injection flow rate may exceed 20 GPM. The correct answer is A. A: Correct -High seal water injection filter dip is an indication of EITHER a clogged filter (Inadequate flow to RCP seals) or high total seal water injection flow rate (> 20 GPM/Pump).

B: Incorrect -Increased dose rate from the filter may be an ALARA issue, but is NOT the initial concern. C: Incorrect -10 GPM IPump is higher than normal, the action is to ensure < 20 GPM/Pump. 0: Incorrect -Total seal injection is normally in excess of 20 GPM. Exam Question Number: 49

Reference:

APP-001-A2; OP-301, Page 14, Step 5.59; SO-021, CVCS, Page 11, Figure 10. KA Statement: Knowledge of the effect that a loss or malfunction of the CVCS will have on the following: RCP seal injection. History: New -Written for HLC-08 NRC exam. 56 ALARM SEAL WTR INJ FILTER HI , AUTOMATIC ACTIONS 1. None Applicable CAUSE 1. Dirty Filter 2. High Seal Water Injection Flow OBSERVATIONS

1. RCP Thermal Barrier (PI-131A, PI-128A and PI-125A) ACTIONS 1. Dispatch an operator to check Seal Water Injection Filter and Seal Injection flow. (PIC-157, FI-124, FI-127, and FI-130 Charging Pump Room) APP-001-A2

2. IF RCP Seal Injection Flow is high, THEN verify RCP Seal Injection Flow is less than 20 gpm to each RCP. \ 3. IF RCP Seal Injection Filter is high AND is NOT caused by high flow, THEN shift filters using OP-301. 4. IF RCP Seal Injection Filters are shifted due to high THEN initiate action to replace the affected filter cartridge.

5. IF RCP Seal Injection flow can NOT be maintained greater than 6 gpm to each RCP, THEN refer to AOP-018 AND ITS SR 3.4.17.1.

DEVICE/SETPOINTS

1. PIC-157 120 psid POSSIBLE PLANT EFFECTS Loss of RCP Seal Injection Flow REFERENCES

1. AOP-018, Reactor Coolant Pump Abnormal Conditions

2. OP-301, Chemical and Volume Control System (CVCS) 3. CWO 8-190628, Sheet 595, Cable A 4. ITSSR3.4.17.1 I APP-001 Rev. 41 Page 5 of 541 5.56 When more than one Charging Pump is operating, only one Charging Pump should be operated in automatic to prevent the Charging Pumps from "hunting" and causing level swings. 5.57 The following starting duty limitations apply to the Charging Pump motors: (ACR 92-325) -Maximum number of starts per hour is 4. -Minimum time between starts is 5 minutes. 5.58 The following starting duty limitations apply to the Boric Acid Transfer Pump motors: (ACR 92-325) -Maximum number of starts per hour is 13. -Minimum time between starts is 3 minutes . . 5.59 Normal Seal Injection flow should be maintained at 8 to 13 gpm, however the minimum Seal Injection flow is 6 gpm and the maximum Seal Injection flow is 20 gpm. (ACR 94-01811)

ITS LCO 3.4.17 requires seal injection flow of 6 gpm to each RCP when in MODES 1, 2, 3, and 4. 5.60 Care should be exercised when rinsing in a Mix Bed or Deborating Demineralizer to prevent from reaching the minimum VCT temperature (60°F). Temperatures less than 60°F can result in low seal injection temperatures, which can cause No. 1 Seal Leakoff Flows to be below normal. 5.61 If the starting limitations stated below are exceeded, Primary Water Pump motor damage can occur due to motor overheating: (REF: ACR 92-325) -Maximum number of starts per hour is 20. -Minimum time between starts is 2 minutes. 5.62 This procedure has been screened in accordance with PLP-037 criteria and determined not applicable (N/A) to PLP-037. 5.63 If additional Charging Pump(s) are needed to combat a casualty, HCV-121 should be opened prior to starting additional Charging Pumps to ensure the Charging Pump Relief Valves will not lift. IOP-301 Rev. 90 Page 14 of 1141 SD-021 CHEMICAL AND VOLUME CONTROL SYSTEM dilute and borate. These modes of operation will be described in the control functions section. The three flowpaths for makeup flow are as follows: Normal Makeup Flow (Automatic) - The makeup system automatically supplies water at the current RCS boron concentration to the charging pump suction as a function of VCT level. Primary makeup water is supplied to the blender by the primary water pumps through valve FCV-114A. Concentrated boric acid is supplied to the blender by the boric acid transfer pumps through valve FCV-113A. The blended solution flows through valve FCV-113B to the charging pump suction header. Emergency makeup from the RWST can be supplied to the charging pump suction header through valve LCV-115B. Boration Flow -Identical to normal makeup flow except that primary makeup is not supplied to the blender. Dilution Flow -Primary makeup water is supplied to the blender by the primary water pumps through valve FCV-1l4A. FCV-1l4B, blender outlet to the VCT opens supplying primary water directly to the VCT. This flowpath provides better control of positive reactivity addition by allowing the primary water to mix with boric acid in the VCT rather than providing primary water directly to the suction of the charging pumps. This flowpath also provides hydrogen addition to the primary makeup water as it enters the top of the VCT. Alternate Dilute -Identical to dilution flow but also supplies primary makeup to the charging pump suction header as well as to the VCT spray nozzle. 2.3.5 Seal Water Injection (Figure 10) eves Total seal water flow rate can be adjusted by charging flow control valve HCV-121, which is normally full open. Manual valves CVC-297 A, B & C, located in the Charging Pump Room, are provided for adjusting the seal water flow for each RCP. The limits on flow are 6 -20 gpm for each RCP, but normal band is 8 -13 gpm. One charging pump operating at minimum speed will provide approximately 18 -24 gpm flow. HCV-121 and/or the seal injection manual control valves must be maintained open enough to pass all of the charging pump flow to prevent lifting a charging pump relief (see CR 95-1752 in the Operating Experience Section of this SD). Flow of seal water is from the discharge of the charging pump through one of two seal water injection filters to just below the lower radial bearing of the RCP. Seal water bypass flow is from the high pressure side of the number I seal to the seal water return header. Seal water return from the low pressure side of the number 1 seal goes to the seal water return header. Page 11 of71 Revision 1 0 INFORMATION USE ONLY fSE,ul-' L ____ -' riW2! I-___ J ,--:----, I SEALI I l.. ___ J SEAL INJECTION CVCS-FIGURE-l 0 SEAl I BYPASS FROM RCPB&C evC-307 HVC-137 FROM lOOP TO lEG lOOP 2 'Q'7 RCDT l r I I EXCESS I.EmOWN 389 I HEAT EXCHANGER -.--J F I SEAl. WATER FILTER TO VCT SEAl. WATER INJECTION FILTERS TORCPB&C FROM CHARGING PUMP DISCH. I IFI ... [ CVCSF10J INFORMATION USE ONLY HLC-08 NRC Written Exam 50. Given the following: -The plant is operating at 100% RTP. -APP-003-C3 PRT HI PRESS illuminates and PRT pressure is still increasing. -PRT level is 70% and stable. -PRT Temperature is 105 of and stable. Which ONE (1) of the following describes the event? A. A PZR PORV or Safety valve is leaking. B. The PRT Nitrogen supply regulator has failed. C. Primary Makeup Water to the PRT is open/leaking. D. RCP Seal Return Relief has lifted and is stuck open. 50 HLC-08 NRC Written Exam 50.007 Al.02 OOllPRT/QUENCH TANKl2/1/2.7/2.9/RO/HIGH/N/A/NEW -20081PZR-008 Given the following: -The plant is operating at 100% RTP. -APP-003-C3 PRT HI PRESS illuminates and PRT pressure is still increasing. -PRT level is 70% and stable .

• PRT Temperature is 105 of and stable. Which ONE (1) of the following describes the event? A. A PZR PORV or Safety valve is leaking. B:I The PRT Nitrogen supply regulator has failed. C. Primary Makeup Water to the PRT is open/leaking.

D. RCP Seal Return Relief has lifted and is stuck open. The correct answer is B. A: Incorrect -No temperature or level increase is indicated. B: Correct -Only pressure is increasing. Nitrogen pressure with no regulation could pressurize the PRT to rupture disk pressure. C: Incorrect -No level change is indicated. D: Incorrect -No level/temperature change indicated. Exam Question Number: 50

Reference:

APP-003-C3; SD-018, Compressed Gas, Pages 19-20, Figure 2. KA Statement: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRTS controls including: Maintaining quench tank pressure. History: New -Written for HLC-08 NRC exam. 57 ALARM PRT HI PRESS AUTOMATIC ACTIONS 1. Not Applicable CAUSE APP-003-C3

1. In leakage from Makeup Water, Pressurizer Relief Valves, Pressurizer Safety Valves, RHR Loop Relief Valves, Letdown Relief Valves, Seal Water Return Relief Valve, SI Test Line Relief Valve, or SI Cold Leg Injection Header Relief Valve

3. Failure of N2 Supply to PRT Opening of Pressurizer Safety or PORV OBSERVATIONS

1. PRT Level (LI-470) 2. PRT Pressure (PI-472) 3. PRT Temperature (TI-471) 4. Pressurizer Safety Valve Line Temperatures (TI-46S, TI-467, TI-469) S. PORV Discharge Line Temperature (TI-463) ACTIONS 1. IF a PZR PORV or Safety fails open while greater than 3S0°F, THEN Refer To Path-1. 2. IF pressure is high, THEN vent the PRT as follows: 1) Open RC-S49, PRT VENT 2) IF required, THEN verify a Waste Gas Compressor starts. 3) WHEN pressure is less than 3 psig, THEN close RC-S49. 3. IF necessary, THEN adjust Nitrogen Regulator to PRT. 4. IF necessary, THEN drain the PRT using OP-103. DEVICE/SETPOINTS

1. PC-472/S psig POSSIBLE PLANT EFFECTS 1. PRT Rupture Disk failure at 100 psig REFERENCES

1. Path-1, EOP Network 2. CW D B-190628, Sheet 461, Cable P 3. OP-103, Pressurizer Relief Tank Control System I APP-003 Rev. 37 Page 25 of 531 SD-018 NITROGEN AND HYDROGEN SYSTEMS ATTACHMENT 10.2 Page 1 of2 High Pressure & Low Pressure N2 Loads LOADS FROM THE LOW PRESSURE SOURCE/SETPOINT NITROGEN SUPPLY Steam Dump Accumulators PCV -1090 175 psig Nitrogen Injection to Condensers A&B Off supply to Steam Dump Accumulators via PRV-ll054 reduced to 50 psig Steam Dump Valves PCV -1090/1 091 S/G PORV s (backup to IA) reduced to 125 psig S/G Wet Layup PCV -1090/1091 reduced to 5 psig via PCV-1019 Resin Storage Tank: PCV -1089 105 psig VCT (via PCV-1l9) RCDT (via PCV-1014 -3psig) PRT (via PCV-473) H2 Recombiner Gas Compressors Bubblers Gas Analyzer WGDTs (via PCV -1046 -15 psig) Gas Strippers "A" & "B" (via PCV-1046-15 psig) Spray Additive Tank (via PCV -1046 -15 psig) CVCS Holdup Tanks (via PCV-1049 -0.65 psig) COMPRESSED GAS Page 19 of 20 INFORMATION USE ONLY BACKUP Bank "G" Cylinders

-PCV -1091 ... 160psig Same as Steam Dump Accumulators Bank: "A" Cylinders -PCV -1043 ... 100 psig Bank "B" Cylinders -PCV -1044 ... 90 psig Revision 7 SD-018 NITROGEN AND HYDROGEN SYSTEMS ATTACHMENT 10.2 Page 2 of2 High Pressure & Low Pressure N2 Loads LOADS FROM THE ruGH PRESSURE SOURCE/SETPOINT NITROGEN SUPPLY Bank "G" Cylinders (Steam Dump Backup) PCV-1091 -160 psig Banks "C", "D", "E" & "F" Cylinders -High Pressure Header PPS Backup Banks "A" & "B" Cylinders High Pressure Header PPS PCV-1809 -100 psig IVSW Backup to: PCV-1043 -100 psig PCV-1044 -90 psig Resin Storage Tank VCT (via PCV-119) RCDT (via PCV-1014 -3psig) PRT (via PCV-473) H2 Recombiner . Gas Compressors Bubblers Gas Analyzer WGDTs (via PCV-1046 -15 psig) Gas Strippers "A" & "B" (via PCV-1046 -15 psig) Spray Additive Tank (via PCV-1046 -15 psig) CVCS Holdup Tanks (via PCV-1049 -0.65 psig) SI Accumulators PCV -937 -between 665 Pressurizer PORV s psig and 685 psig COMPRESSED GAS Page 20 of 20 INFORMATION USE ONLY BACKUP Revision 7 STEAM ... T DUMPS S/G PORVs S/G WET ........ I-__ LAYUP 1097 175# NITROGEN SYSTEM GAS-FIGURE-2 Condenser

c Ii" A&B E w w C -<> t c li! ::J VI 0...: ....... HP*AMBIENT AIR VAPORIZER ii2 "'" < U OZ 1i;.1i;

... < 0 ... '--AIA_ N2 TO BUBBLER eves HOLD-UP 0.65 TANKS __ ________ -L __ 160# PCV.1091 IVSW 1049 PPS BANK G BANK F BANK E BANK 0 BANK e 1\ I STEAM PPS BACKUP PPS BACKUP DUMP NS-17 TRUCK FILL II-I ...... 1---' BACKUP \ I LP RESERVE gasf02 INFORMATION USE ONLY BANKB HLC-08 NRC Written Exam 51. Which ONE (1) of the following describes the design features which act to limit S/G blowdown rate for a Steam Break accident? A. Swirl Vane Separators in the upper S/G internals. B. A set of 7 venturis in the S/G outlet nozzles. C. Flow venturis in each steam line, located at the operating floor. D. MSIVs are required to close within 5 seconds. 51 HLC-08 NRC Written Exam 51.039 K1.01 OOlIMAINIREHEAT STEAMJ2/113.1I3.2fROILOWININNEW -2008/SGS-004 Which ONE (1) of the following describes the design features which act to limit S/G blowdown rate for a Steam Break accident? A. Swirl Vane Separators in the upper S/G internals. A set of 7 venturis in the S/G outlet nozzles. C. Flow venturis in each steam line, located at the operating floor. D. MSIVs are required to close within 5 seconds. The correct answer is B. A: Incorrect -Swirl vanes are designed to impart radial motion to the steam to remove moisture. B: Correct -S/G Steam nozzles are provided with flow limiting devices to restrict the steam flow resulting from a Steam Line Break. C: Incorrect -Flow venturis in the lines at floor level produce a small delta P to the steam as a means of measuring flow. Pressure drop (and the flow limiting effect) is designed to be small for efficiency reasons. 0: Incorrect -MSIV closure time would have no effect on UPSTREAM breaks. Exam Question Number: 51

Reference:

SO-025, Main Steam, Page 9, Figure 1; SO-048, S/G, Page 8, Figures 4 and 8. KA Statement: Knowledge of the physical connections and/or cause-effect relationships between the MRSS and the following systems: S/G. History: New -Written for HLC-08 NRC Exam. 58 SD-025 MAIN STEAM SYSTEM The three steam lines connect to a 72-inch header. The 72-inch header then divides into two headers (generally referred to as the North and South Headers). The 72-inch header has the following:

1. Drains 2. Vent valves (used to ensure steam system remains depressurized during outage) 3. Taps for header pressure instrumentation The North and South header supply steam seals, condenser steam dumps, heating steam to MSRs, and the turbine. These headers have drain taps to remove the moisture from the steam line continuously and can be drained to the atmosphere when warming steam lines. These headers feed the left and right side of the turbine. The steam is exhausted below the LP turbines directly to the Main Condensers.

3.0 COMPONENT DESCRIPTION 3.1 Flow Nozzles Steam Flow Limiter Number 3 MSS Type Size Venturi Material Steam Flow Nozzles Number Type Work pressure Work Temperature Pipe Size Cone and Throat Material Throat Diameter 7 venturies in a machined disc Maximum throat area of the 7 venturis is 1.4 sq. ft. Inconel 600 casting 3 Flow Tube, Fabricated insert 1085 psig 600°F 26" 304 Stainless 16.469 + 0.010 Each main steam line contains two types of steam flow nozzles. One type is a steam flow limiting device located within the generator shell steam outlet nozzle. The other fl2.w nozzle (venturi) is located further downstream in steam pipeline just above the containment operating floor. Page 90f44 Revision 7 INFORMATION USE ONLY v PT @ nAn S/G SYSTEM DIAGRAM (S/G TO 72" HEADER) MSS-FIGURE-l (Rev 0) ATMOS ATMOS ATMOS ATMOS ATMOS SAFETIES MS SAMPLE STEAM DRIVEN AUX. FEEDWATER PUMP 72" HEADER NORTH { r, ATMOS SOUTH D jmSSf01] SD-048 STEAM GENERATOR SYSTEM the ReS during accident conditions, as required. This ensures the ability to achieve safe shutdown of the reactor and permits natural circulation of the ReS when required. The S/G limits steam flow for any downstream Main Steam Line Break (MSLB). The S/G steam nozzles are provided with flow limiting devices to restrict the steam flow resulting from a steam line break. The flow restrictors protect the containment from exceeding its design pressure and prevent excessive cooldown rate of the ReS by limiting the steam flow that can result from a MSLB. Normal Operating Functions The S/Gs convert the heat generated in the ReS into steam as required for the turbine and auxiliaries during all modes of Main Steam System operation. The design of the S/Gs includes primary and secondary provisions for the removal of moisture to preclude carryover in excess of 0.25 percent of the steam flow. The first stage of moisture separation occurs in the swirl vanes and the second stage in the contoured vane banks. 2.3 System Flow Paths (Figure 4) SG Primary Side Reactor coolant enters the inlet side of the channel head at the bottom of the S/G through the inlet nozzle, flows through the U-tubes to the outlet side of the channel head and leaves the generator through another bottom nozzle. It then passes out of the S/G to the intermediate leg. The coolant leaves the S/G and flows to the Rep suction through the intermediate leg. Secondary Side Feedwater to the S/G enters just above the top of the U-tubes through a feedwater ring. The water flows downward through an annulus between the tube wrapper and the shell and then upward through the tube bundle where part of it is converted to steam. The steam-water mixture from the tube bundle passes through a steam swirl vane assembly which imparts a centrifugal motion to the mixture and separates the water particles from the steam. The water spills over the edge of the swirl vane housing and combines with the feedwater for another pass through the tube bundle. The steam rises through additional separators which limit the moisture content of the steam to one fourth of one percent or less under all design load conditions. Page 8 of28 Revision 6 INFORMATION USE ONLY WATER LEVEL WIDE RANGE NARROW RANGE 100% 1_ 90'11 80'11 70% 90'11 60'11 50% .0'11 80'11 30'11 ZO'Ii 10'1i 0% 70'11 60'11 .0'11 30% ZO'll 10'11 0'11 .... _-...... _., ... Steam Generator Diagram SG-FIGURE-4 STEAM WITH MOSITUItE SECONDARY SEPARTOR BOTTOM DECK PlATE SECONDARY /MNWAY NORMAL OPERAnNG WATER LEVEL (APPROX) THREE (3) PRIMAAY SEPARATOR SWIRL VANE BARRELS PRIMARY SEPARATOR DOWNCOMER ---fl--it BARREL TYP. 3 PLS. fEEDWATER DISTRIBUTION RING WITH J-NOZZLES WRAPPER SECONDARY SEPARATOR. PERFORATED PlATE (TYP. 8 PLACES) SECONDARY SEPARATOR ............

---

VANES (TYP. 8 PLACES) SECONDAR.Y SEPARATOR 5" SCH .0 DRAIN PIPES (TYP. 16 PLACES)

RANGE LEVel UPPER TAP CENTRAL DRAIN __ ------STEAM VENT (lYP.3 PlACES) NARROW RANGE LEVEL LOWER TAPS ' ... JIM!-__ BAAS TRANSITION CONE DOWNCOMER ANNULUS QUATREFOIL TUBE SUPPORT PlATE STEAM WATER MIXTURE WATER ENTERS HEATING AND SOILING SECTION STAY ROD (INSIDE SPACER PIPE) PERFORATED BOTTOM SPACER PIPE BLOWDOWN PIPE 6'HANOWAY FLOW DISTRIBUTION BAFFLE SIX (6) TUBELANE FLOW WIDE RANGE LEVEL BLOCKING PLATES LOWER TAP 6"HANDHOLE BOTTOM BLOWDOWN DIVIDER PLATE DIVIDER PLATE PRIMARY MANWAY NOZZLE DRAIN MANWAY DRAIN sgro4 STEAM FLOW LIMITER SG-FIGURE-8 sgfOB 1.12 -(REF.) HLC-08 NRC Written Exam 52. Given the following: -A Loss of Instrument Bus 1 has occurred. -The crew is performing actions contained in AOP-024, LOSS OF INSTRUMENT BUS. -The Pressurizer Pressure Controller (PC-444J) AUTO light illuminates for approximately 15 seconds, then extinguishes, and then the MANUAL light is illuminated. Which ONE (1) of the following describes the event in progress and the associated action, if any? A. Instrument Bus 1 power is NOT restored. Operate Pressurizer Heaters and Spray valves manually. B. Instrument Bus 1 power is restored. PC-444J AUTO has failed, continue to operate in MANUAL. C. Instrument Bus 1 power is restored. Restore PC-444J to AUTO control. D. Instrument Bus 1 power is NOT restored. Operate PC-444J in MANUAL. 52 HLC-08 NRC Written Exam 52. 062 A3.05 OOI/AC ELECTRICAL DIST/2/1/3.5/3.6/RO/HIGH/N/AJRNP AUDIT -2001lAOP-024-004 Given the following: -A Loss of Instrument Bus 1 has occurred. -The crew is performing actions contained in AOP-024, LOSS OF INSTRUMENT BUS. -The Pressurizer Pressure Controller (PC-444J) AUTO light illuminates for approximately 15 seconds, then extinguishes, and then the MANUAL light is illuminated. Which ONE (1) of the following describes the event in progress and the associated action, if any? A. Instrument Bus 1 power is NOT restored. Operate Pressurizer Heaters and Spray valves manually. B. Instrument Bus 1 power is restored. PC-444J AUTO has failed, continue to operate in MANUAL. CY-Instrument Bus 1 power is restored. Restore PC-444J to AUTO control. D. Instrument Bus 1 power is NOT restored. Operate PC-444J in MANUAL. The correct answer is C. A: Incorrect -The controller has NOT failed, it indicated it was returning to the normal source of power. B: Incorrect -Power has been restored to PC-444J which operated through its normal power restoration cycle. The controller does NOT automatically swap to the Auto mode. C: Correct -When power is restored to PC-444J, the Auto indicator will flash indicating power is returned to the normal mode. The controller will stay in Manual control. Operator action is required lAW AOP-024 to restore controller to Auto. D: Incorrect -Power is restored instead of momentarily restored. Power restoration will leave the controller in MANUAL. Controller can be operated in MANUAL or AUTOMATIC while on alternate power source. Exam Question Number: 52

Reference:

AOP-024, Page 5; OP-001, Pages 23-25. KA Statement: Ability to monitor automatic operation of the ac distribution system, including: Safety-related indicators and controls. History: Direct from Bank. 59 AOP-024 LOSS OF INSTRUMENT BUS CONTINUOUS USE ATTACHMENT 1 EXTENDED LOSS OF INSTRUMENT BUS 1 (AND 6) CPage 1 of 5) The following control functions/indications will be lost until Instrument Bus 1 and 6 are restored: Steam Dump Tavg mode of operation FRV A Automatic control FRV Bypass Valve B CFCV-489) PCV-145 Controller Clocks up) PZR Heaters PZR Pressure Controller. PC-444J Clocks up) PZR Level Controller. LM-459G Clocks up) PZR PORV. PCV-455C PZR Spray Valve Controller. PC-444G Clocks up) PZR Spray Valve Controller. PC-444H Clocks up) PZR Safety Acoustic Monitors Charging Pump Controller A. SC-151A Clocks up) S/G B PORV FCV-113A Controller Clocks up) VCT Level Controller. LC-112 Clocks up) Excess Letdown Temperature Channel TI-139 Rev. Page 1. Continue to operate PCV-464B. Steam Dump Controller. in Manual Pressure mode. 2. Continue to operate the FRV A in MAN mode. 29 24 3. Contact Operations Staff for availability of a dedicated FRV watch. 4. IF CHARGING PUMP A is in service. THEN perform the following:

a. IF necessary.

THEN start CHARGING PUMP B OR C. b. Stop CHARGING PUMP A. 5. Control PZR level by manual control of Charging Pump speed. of 84 CONTINUOUS USE Section 8.4.3 Page 1 of 4 NOTE: This section has been screened lAW PLP-037 criteria and determined to be a Case Three activity. No additional management involvement is required beyond that routinely provided by first line supervision. ITS LeO 3.8.7 and LCO 3.S.S should be referenced for instrument bus operability requirements. Transferring Instrument Bus 1 Power Supply{ TC "Transferring Instrument Bus 1 Power Supply" \f C \1 "3" } IOP-001 1. Initial Conditions

a. This revision has been verified to be the latest revision available.

Date b. IF transferring to the alternate power supply, THEN VERIFY MCC-S is energized AND no other Instrument Bus is being supplied by MCC-S. c. IF transferring to the normal power supply, THEN VERIFY MCC-5 is energized.

d. CHECK that Reactor Trip Breakers are OPEN OR Reactor Power is greater than P-10. e. PERFORM The Following:

1) EVALUATE the affect of a loss of Instrument Bus 1 on illuminated bistables in channels II, III, and IV 2) IF the loss of Instrument Bus 1 concurrent with illuminated bistables in channels II, III or IV will cause a transient, THEN CLEAR the affected bistables prior to continuing.

Rev. 25 Page 23 of 431 Section 8.4.3 Page 2 of 4 8.4.3.1 (Continued) NOTE: Performance of this section will result in the following: RCS Letdown Isolation Momentary loss of PZR Heaters Momentary loss of all PZR Pressure and Level Controllers

f. REVIEW EDP-008 for additional loads that will be lost. 2. Instructions NOTE: Loss of Instrument Bus 1 will result in isolating letdown and tripping PZR heaters from the PZR Low Level Relay (LC-459C1-X).

This can only be avoided by manually overriding the relay the entire time it is deenergized. IOP-001 a. ISOLATE Letdown using OP-301. b. VERIFY PZR Heaters are OFF. c. PLACE FRV "A" controller in (FCV-478) in MAN. d. IF the plant is in Mode 1, THEN verify the following: -Rod Control in Manual. -Turbine Control in Manual. Rev. 25 Page 24 of 431 Section 8.4.3 Page 3 of 4 8.4.3.2 (Continued) CAUTION Performance of the following step will result in a momentary loss of power to Instrument Busses 1 and 6. e. PLACE the INST BUS 1 PWR XFER SW to the desired position. NORM/ ALT (Circle one) f. RESET the Dropped Rod Alarm by momentarily placing the Dropped Rod Mode Selector switch for N-41 to RESET. NOTE: When a Manual/Auto control station is reenergized, 15 to 20 sec. is needed for the AUTO light to go out and the Manual/Auto station to revert to the manual mode. IOP-001 g: RESTORE affected controllers on the RTGB to AUTO. h. RESTORE Letdown using OP-301. i. IF desired, THEN PLACE Rod Control in AUTO. j. IF desired, THEN PLACE Turbine Control in AUTO. Rev. 25 Page 25 of 431 QUESTIONS REPORT 1. 057 AA2.14 001 for AUDIT (;2 00 t) A Loss of Instrument Bus 1 has occurred. The crew is performing actions contained in AOP-024, Loss of Instrument Bus. The Pressurizer Pressure Controller PC -444J, AUTO light comes on for approximately 15 seconds, then goes out, and then the MANUAL light is illuminated. Which ONE (1) of the following describes the event in progress and the associated action? A. PC-444J AUTO controller has failed. Operate Pressurizer Pressure control in Manual as directed by AOP-019, Malfunction of RCS Pressure Control. B. Power is restored to PC-444J. The controller has failed to swap to AUTO. Operate Pressurizer Pressure control in Manual as directed by AOP-019, Malfunction of RCS Pressure Control. Power is restored to PC-444J. Restore to Automatic control as directed by AOP-024. D. Power was momentarily restored to PC-444J. When the controller AUTO light is illuminated again, verify Auto control by adjusting the controller in accordance with AOP-024. C is correct SRO Question 080 Tier 1 Group 1 KIA Importance Rating -SRO 3.6 Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus: That substitute power sources have come on line on a loss of vital AC. Reference(s) -VAC SO. AOP-024 Proposed References to be provided to applicants during examination -None Learning Objective -Question Source -New Question History -Question Cognitive Level -Comprehension 10 CFR Part 55 Content -43 Comments -Category 1: Category 3: Category 5: Category 7: Tuesday, June 10,200812:35:46 PM Category 2: Category 4: Category 6: Category 8: 1 Rev. 29 AOP-024 LOSS OF INSTRUMENT BUS INSTRUCTIONS

6. Determine Failed Instrument Bus (IB) From Any Of The Following:

• Available indications
• Table Below Page 5 of RESPONSE NOT OBTAINED Inst Indication To Check Bus 1 FR-478, "A" SIG Level 2 FR-488, "B" SIG Level 3 FR-498, "C" S/G Level 4 TR-408, Tavg 6 LI-459A, PZR Level 7 LI-460, PZR Level 8 LI -461, PZR Level 9 LI-462, PZR Level Affected controllers may be returned to AUTO anytime the Instrument Bus power is regained.

Controllers will take 15 to 20 sec. to return to manual after reenergizing.

• 7. Check Emergency Busses E-l AND E-2 -ENERGIZED FROM THE 4160V BUSSES IF an Emergency Bus has deenergized, THEN check the EDG STARTS and its output breaker CLOSES. IF the EDG has NOT auto-started, THEN initiate a start from the RTGB (will involve a 4.5 minutes delay). 84 HLC-08 NRC Written Exam 53. Given the following:

-Reactor Trip and Safety Injection have occurred. -The following alarms were received in the Control Room: -APP-004-A1, S/G A STM LINE HI DELTA P SFGRDITRIP -APP-004-A5, S/G A LO LVL & STM > FWF TRIP -APP-006-A2, S/G A STM > FW FLOW -APP-006-E5, STM LINE LO PRESS -The crew has completed Supplement G, STEAM GENERATOR ISOLATION. Which ONE (1) of the following describes the Main Steam System component(s) controlling RCS Heat Removal? A. Condenser Steam Dumps from S/Gs "A", "B", and "C". B. S/G PORVs "A", "B", and "C". C. S/G "B" and "C" Main Steam Safety Valves. D. Condenser Steam Dumps from S/Gs "B" and "C". 53 HLC-08 NRC'Written Exam 53. 039 K3.06 OOllMAINIREHEAT STEAM/2/1I2.8/3.1IRO/HIGH/N/NRNP BAN:K/EPP-II-005 Given the following: -Reactor Trip and Safety Injection have occurred. -The following alarms were received in the Control Room: -APP-004-A 1, S/G A STM LINE HI DELTA P SFGRDITRIP -APP-004-A5, S/G A LO LVL & STM > FWF TRIP -APP-006-A2, S/G A STM > FW FLOW -APP-006-E5, STM LINE LO PRESS -The crew has completed Supplement G, STEAM GENERATOR ISOLATION. Which ONE (1) of the following describes the Main Steam System component(s) controlling RCS Heat Removal? A. Condenser Steam Dumps from S/Gs "A", "S", and "C". B. S/G PORVs "A", "S", and "C". C. S/G "s" and "c" Main Steam Safety Valves. Condenser Steam Dumps from S/Gs "s" and "C". The correct answer is D. A: Incorrect -Steam Dump NOT available from S/G "A". S/G "A" isolated per Supplement G. S: Incorrect -S/G PORV is available from S/Gs "s" and "C". S/G "A" is isolated lAW Supplement G. C: Incorrect -Steam Dumps are available to condenser from S/Gs "s" and "c" and Steam Line PORVs "S" and "C". Normal control system setpoints will prevent Main Steam Line Safety valves from lifting. D: Correct -S/G "A" is isolated lAW Supplement G. S/Gs "S" and "c" are aligned to dump steam to condenser. Exam Question Number: 53

Reference:

Supplement G, Page 37; SD-031, Steam Dump, Pages 8, 9 and 19. KA Statement: Knowledge of the effect that a loss or malfunction of the MRSS will have on the following: SDS. History: Direct from Sank. 60 Rev. 35 EPP-Supplements SUPPLEMENTS Page 37 of 89 INSTRUCTIONS RESPONSE NOT OBTAINED CONTINUOUS USE Supplement G Steam Generator Isolation (Page 1 of 12) 1. Go To Appropriate Step From Following Table: S/G TO BE ISOLATED STEP S/G A 2 S/G B 18 S/G C 34 2. Check S/G A -FAULTED --3. Verify Vl-3A. MSIV -CLOSED --4. Verify MS-353A. MSIV Vl-3A BYP -CLOSED WHEN Tavg less than 547°F. THEN perform Steps 3 AND 4. WHEN S/G A level is greater than 8% [18%J. THEN observe the NOTE prior to Step 5 and perform Steps 5 though 9. Verify ruptured STEAM LINE PORV setpoint at 1035 psig using Status Board. Go To Step 11. Local operation of the FRV and B/P valves below is via reverse acting handwheels. -5. Verify FRV A -CLOSED --6. Verify FRV A BYP -CLOSED SD-031 STEAM DUMP SYSTEM the desired steam pressure as set on the steam dump controller and modulates steam dump valves open as necessary to maintain steam pressure at setpoint. Tavg Mode of control has basically two sub-modes: load rejection and turbine trip. The function of either of these modes is to control Tavg to some desired value set by the operator. If the plant were to experience a load rejection there would initially exist a power mismatch between the reactor and turbine until the rods could be inserted to lower reactor power. This mismatch would cause Tavg to increase, which will cause the rod control system to insert rods to lower the temperature. If the load rejection is larger than rod control is designed to handle, the steam dumps would open to give an artificial load to limit the Tavg.rise. The amount of modulation open of the steam dump valves is controlled by the deviation of actual Tavg (median Tavg is used) from program Tavg (Tret). If the deviation exceeds a preset value another signal is sent to the steam dump valves to "pop" open. With the turbine at a lower power level and impulse pressure, the program value for Tref will be lower than before the load rejection. As the rod control system inserts the rods, lowering Tavg, the deviation between actual and program temperature decreases and the steam dump valves modulate shut. The Turbine Trip submode basically performs the same as load rejection mode. When a turbine trip occurs, a power mismatch occurs between the reactor and the secondary system, even if a reactor trip occurs. Power in the reactor does not immediately go to zero, so heat must be removed from the RCS. Since turbine impulse pressure is zero after a trip, Tref goes to its minimum value. If sufficient deviation from no load Tavg is sensed, the steam dump valves will modulate open or "pop" open if the deviation exceeds a preset value. (SD-Figure-2) The system has to meet three "permissives" to be able to modulate the condenser steam dump valves: a circulating water pump must be running, adequate condenser vacuum must be available (> 19.7"Hg Vac; i.e. APP-008-B6, CONDENSER LO VACUUM TURB TRIP, extinguished) and the system must be turned on. In addition to the permissives being met, the steam dump system must be "armed" to allow dump valves to open. In the steam pressure mode of control, the system is armed when Steam Pressure Mode is selected and the steam dlimp valves can modulate as necessary to control main steam header pressure as long as the permissives are met. In the Tavg mode of control further stimulus is required to arm the steam dumps. If a load rejection is sensed, as determined by turbine impulse pressure decreasing by a Steam Dumps Page 8 of 26 Revision 7 INFORMATION USE ONLY SD-031 STEAM DUMP SYSTEM prescribed amount in a preset time, the load rejection submode will "arm" and actuates to help control Tavg along the Tavg program. If a turbine trip occurs, the turbine trip submode "arms" and actuates to reduce Tavg to the no-load value. Since PT-447 arms the Steam Dumps and PT-446 programs Tref for the Steam Dumps, the failure of a single instrument will not cause the Steam Dumps to open when in the Tavg mode (see SD-Figure-6). (SD-Figure-3) The term "arm", when used with the steam dumps, refers to nitrogen being aligned to the steam dump positioner and the high temperature trip open 3-way valves. When nitrogen is aligned to the positioner, the output of the different steam dump controllers can then control the positioner to allow the nitrogen to flow to the diaphragm of the valve and modulate the valve based on the temperature deviation. If the "trip open" signal is present, the 3-way valve can*put full nitrogen pressure to the booster, which in turn permits full nitrogen pressure to pass and be applied to the diaphragm of the valves, which causes the valves to pop open. (SD-Figure-4 ) The steam dump valves, when actuated by the Tavg deviation from Tref in the load rejection submode, operate in banks. The fIrst bank arms on a 15 % load rejection and opens at"a deviation of 5°F and is full open at 11.6°F. The second bank arms on a 35% load rejection and begins to open at 11.6°F deviation and is fully open at 15.9°F. The 4th bank, S/G PORVs, is modulated by the SDCS on a 50% load rejection and starts to open at 15.9°F and is fully open at 20.5°F deviation. The same staggered operation occurs in the turoine trip submode with slightly different setpoints, except that Bank 4, S/G PORVs, will not be controlled by the steam dump system. An interlock exists which turns off condenser steam dumps when Tavg reduces to less than 543°F. This interlock prevents the steam dump actuation from causing an overcooling event. The interlock can be bypassed once Tavg is less than 543°F with a spring-loaded switch on the RTGB, which allows bank 1 steam dump valves to be turned back on. This allows steam dumps to be used to perform a plant cooldown after defInite operator action is performed to confIrm that a reduction in Tavg below 543°F is warranted. Once the plant has been cooled to 450°F during a controlled cooldown (GP-007), the other bank of condenser dump valves can be used after I&C installs jumpers/lifts leads in Miscellaneous Relay Rack 50. Steam Dwnps Page 9 of 26 Revision 7 INFORMATION USE ONLY SD-031 STEAM DUMP SYSTEM 6.0 SYSTEM OPERATION 6.1 Normal Operation The operation of the S/G PORVs and steam dumps are directed by the General Procedures. They can be used for plant cooldown or to maintain RCS temperature at a specified value by maintaining an artificial load. Use of the steam dumps is preferred, if available, since secondary coolant is conserved. During a plant startup, Steam Dumps are initially in the Steam Pressure Mode with the potentiometer set at 7.17 (1005 psig), corresponding to the no-load Tavg value of 547°P. As power approaches 15 % (Note that it is not possible to arm the steam dump system in the Tavg mode if you are less than 15 % load), Steam Dumps are transferred to the Tavg Mode by the following sequence: When Steam Dumps indicate closed, depress MAN on PC-464B, Position the STEAM DUMP MODE Selector to RESET and then to T-AVG, Verify the Steam Dump Valves Remain CLOSED, Depress AUTO on PC-464B, Adjust PC-464B STEAM HEADER PRESS controller to a potentiometer setting of 7.28 (1020 psig). During a plant shutdown when between 10 and 15% power, the steam dumps are shifted to the Steam Pressure Mode and the controller is adjusted to a potentiometer setting of 7.17 (1005 psig). When a plant cool down is commenced, the steam dumps (or S/G PORVs) are used to control the plant cooldown until RHR is placed in service. When Tavg is less than 543°P, the low Tavg interlock must be bypassed, allowing the cooldown to be performed with steam dump bank 1. After the plant has reached 450 o P, condenser steam dump bank 2 can be used to help with the cooldown by I&C installing jumpers and lifting leads in the Miscellaneous Relay Rack (MRR-50). This temporary feature is removed once the plant reaches MODE 5 6.2 Abnormal Operation 6.2.1 Secondary Load Rejection During a secondary load rejection or turbine run back, the steam dump system will actuate to provide an artificial load on the RCS and, in conjunction with rod control, reduce Tavg to within 5°P of Tref, if the SDCS is in the Tavg mode. The five steam dump valves and the S/G PORVs can be operated from the steam dump controls during this evolution. The magnitude of the sudden reduction of turbine first stage pressure will determine which valves are armed: > 15<35, Bank 1; >35, Bank Steam Dumps Page 190[26 Revision 7 INFORMATION USE ONLY QUESTIONS REP9RT ) for AUDIT (;'00/ 1. 040 AK2.01 001111/111/ The Reactor Trip and Safety Injection have occurred. The following alarms were received in the Control Room:

• APP-004-A1, S/G A HI STM LINE HI DELTA P SFGRDITRIP
• APP-004-AS, S/G A LO LVL & STM > FWF TRIP
• APP-006-A2, S/G A STM > FW FLOW
• APP-006-ES, STM LINE LO PRESS The crew has completed Supplement G, Steam generator Isolation.

Which ONE (1) of the following describes the Main Steam System component(s) controlling RCS Heat Removal? A. Condenser Steam Dumps from"A", "B", and "C" SGs. B. "A", "B", and "C" SG PORVs. C. "B" and "C" SG main Steam Safety Valves. Dy' Condenser Steam Dumps from liB" and "C" SG. A. Incorrect. Steam Dump not available from A SG. A S/G isolated per Supplement "G". B. Incorrect. Steam Dump available from condenser. C. Incorrect. Steam Dump available from condenser. D. Correct. Common Question 048 Tier 1 Group 1 KIA Importance Rating -RO 2.6 / SRO 2.6 Knowledge of the interrelations between the Steam Line Rupture and the following: Valves. Reference(s) -Supplement G, SD-031 pg. 18; No LP for Supplements found. Proposed References to be provided to applicants during examination -NONE Learning Objective -Question Source -Bank Question History -Question Cognitive Level -Comprehension 10 CFR Part 55 Content -55.41 Comments -Category 1: Category 3: Category 5: Category 7: Tuesday, June 10, 200812:56:17 PM Category 2: Category 4: Category 6: Category 8: 1 HLC-08 NRC Written Exam 54. Given the following: -A Large Break LOCA has occurred with all equipment operating as designed. -Containment pressure peaked at 38 PSIG and is now at 11 PSIG. -The crew has transitioned to EPP-9, TRANSFER TO COLD LEG RECIRCULATION, due to RWST level < 27%. -Ability to recirculate from the Containment ECCS Sump exists. Which ONE (1) of the following describes the manipulation(s) necessary lAW EPP-9, to allow stopping a CV Spray Pump without locking out the pump? A. Reset SI. B. Reset SI, Phase A, and Phase B. C. Place Containment Spray Key Switch to the OVRD/RESET Position. D. Reset SI and momentarily place Containment Spray Key Switch to the OVRD/RESET Position. 54 HLC-08 NRC Written Exam 54.026 A2.04 OOl/CTMT SPRAY/2/113.9/4.2IROIHIGHlNINRNP BANKlCSS-007 Given the following: -A Large Break LOCA has occurred with all equipment operating as designed. -Containment pressure peaked at 38 PSIG and is now at 11 PSIG. -The crew has transitioned to EPP-9, TRANSFER TO COLD LEG RECIRCULATION, due to RWST level < 27%. -Ability to recirculate from the Containment ECCS Sump exists. Which ONE (1) of the following describes the manipulation(s) necessary lAW EPP-9, to allow stopping a CV Spray Pump without locking out the pump? A. Reset SI. B. Reset SI, Phase A, and Phase B. Place Containment Spray Key Switch to the OVRD/RESET Position. D. Reset SI and momentarily place Containment Spray Key Switch to the OVRD/RESET Position. The correct answer is C. A: Incorrect -Resetting SI does NOT have any effect on CV Spray. B: Incorrect -CV Spray must be reset prior to resetting Phase B. C: Correct -CV pressure remains higher than 10 PSIG. Signal must be overridden to allow CV Spray Pump to be stopped and NOT locked out. 0: Incorrect -Reset 81 does NOT have any effect on reset of CV Spray. Unless the Key Switch remains in the OVRD position CV Spray Pump will be locked out due to CV pressure remaining> 10 PSIG. Exam Question Number: 54

Reference:

EPP-9, Step 3; SD-006, ESFAS, Page 22, Figures 4 and 5. KA Statement: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CSS controls including: Containment Pressure. History: Direct from Bank. 61 Rev. 31 EPP-9 TRANSFER TO COLD LEG RECIRCULATION Page 4 of INSTRUCTIONS RESPONSE NOT OBTAINED ************************************************************************** CAUTION Steps 1 Through 24 must be performed without delay to accomplish switchover prior to RWST level reaching 9%. **************************************************************************

• Foldouts are NOT applicable during the performance of this procedure.
• Functional Restoration Procedures are NOT applicable until after Step 42.
• 1. Check Capability To Establish Recirculation

-EXISTS

• Establishment of 354 inches in the CV Sump is possible
• Establishment of at least one flow path from the CV Sump to the RCS is possible 2. Reset SAFETY INJECTION

--? 3 . Place The CONTAINMENT SPRAY Key Switch To The OVRD/RESET Position 4. Verify RCPs -ALL STOPPED Go To EPP-15. Loss Of Emergency Coolant Recirculation. 40 SD-006 ENGINEERED SAFETY FEATURES SYSTEM 3. 880A, 880B, 880C, 880D, 845A, and 845B. This will deliver borated water with sodium hydroxide to the CV atmosphere to depressurize and remove free iodine. A Containment Spray signal can be reset/overridden after actuation if it becomes necessary to stop or realign equipment actuated by the Containment Spray signal. Resetting/Overriding the signal will not terminate Containment spray or cause any component actuated by the Containment Spray signal to change state. Once the signal is overridden, no further automatic Containment Spray actuations will occur until all automatic actuation signals have cleared. A key operated CV SPRAY RESET switch on the RTGB is used to reset a CV Spray signal. Operation of the switch from the NORMAL position will actuate an annunciator on APP-002-Cl; FEEDWATER ISO/CV SPRAY OVRD/RESET. Until the Containment Spray signal is reset/overridden, any Containment Spray actuated equipment stopped from the RTGB cannot be restarted without removing and reinstalling that equipment's control power fuses. This is due to the anti-pump feature of the equipment breakers. The Phase "B" Containment Isolation must be reset individually after the Containment Spray signal is cleared or overridden. Phase "B" Containment Isolation is reset from a pushbutton on the RTGB. This is normally required in the EOP network when < 4 psig in the Containment. This allows stopping components and restoring Phase "B" isolation valves. Phase "B" Containment Isolation -This signal will further isolate the containment by shutting containment isolation valves as follows:

• CC-716A CC to R.C.P. "A", "B", "c" and C.R.D. Coollsol
• CC-716B CC to R.C.P. "A", "B", "C" and C.R.D. Coollsol
• CVC-381 RCP Seal Water Return
• CC-735 CC from R.C.P. "A", "B", "C" Thermal Barrier Iso1.
• CC-730 CC from R.C.P. "A", "B", "C" Oil Cool Iso1.
• FCV-626 CC From R.C.P. "A", "B", "C" Thermal Barrier Isol. 6.3 Station Blackout ESF Occurs on a sensed loss of AC power ( < 328V or degraded grid setpoint of < 430V for ten seconds) to the emergency busses (El and/or E2). Degraded Grid protection is to open the normal supply breaker to the emergency bus. The UV relays (2) will now see a complete loss of voltage and start the blackout sequence.

Note that 125VDC power must be present for the blackout sequencer to operate. Page 22 of 40 Revision 10 INFORMATION USE ONLY ENGINEERING SAFETY FEATURES ESF-FIGURE-4 CONTAINMENT CONlROL ROOM 1WIl0AC11\lI1Y AA£A NMENT' PRESSURE .. I_ tMNUAl.. AClUAilON FROM CONTROL BOARD .. t .. DE'I'ECTORS .. I.MONITOR .. ] :rEO CONTAINMENT lSOIAllON SIGNAL (PlIASt MB) AN/) '£D BY CONTAINI.IENT V£HTILAllON ISOlATION AA£ AU. D IN (lATCHED). SO 1KA.T lOSS OF' mE ACTUATION SIGNAl ESE COI.II'OtlENlS 10 RETURN TO THE POSllION H£t.() Nt OF THE ACTIIAlION SIGHAI.. " PC-95IB ARE fEO BY THE DIESEL AUTnl.lAlIr .... IIY S1EAM UHE ISOlAlIOH 3 , _. Ii .. 9. COMPONENlS ACTUATEO BY SI SEAI..EO IN (LATCIltD) SO SIGNAL WIU. Nar CAUSE 10 THE POSIllON HELD PRIOR 10 THE NM!.Ilr Of" 111£ ACTUAllON SIGNAL. 'UJ COI.IPONENlS AClUAltD BY SI SIGHAI.. CONT. \IFtlI'. ANn I I.-.-I I t r SAFEGUARD ACTUATION SIGNALS ESF-FIGURE-5 I': 5TFAM GENERATOR -,- "1----. 0/ HI HI PRESS. OR HIGH STEAM UNE' I..OW PRESSURIZER . HIGH STFAU UNE flOW DIFFERENTIAl.. I PRESSURE COINCIOBIT WITH I..OW ""''''URE STE"AM UNE PRESSURE ......

• OR LOW TAVG (SHEET 7) (SHEET 6) i<NOTE 7) i! 1-2 MIN. CONTAINMENT

'-----------tl S.I. (AUTO &: MAN.) , / v STE'AM UNE ISOLATION INFORMATION USE ONLY QUESTIONS REPORT 1 . 026 Al.04 001/111/111 for AUDIT (0 I Pr,: ... f}ueJrr) Given the following conditions:

• A LBLOCA has occurred with all equipment operating as designed.
• Containment pressure peaked at 38 psig and is now at 11 psig.
• The crew has transitioned to EPP-9, TRANSFER TO COLD LEG RECIRCULATION, due to RWST <27%.
• Ability to recirculate from the Containment Sump exists. Which ONE (1) of the following describes the manipulation(s) necessary to allow stopping a CV Spray Pump without locking out the pump? A. Reset SI B. Reset 81, Phase A, and Phase B CY' Place Containment Spray Key Switch to the OVRD/RESET Position D. Place Containment Spray Key Switch to the OVRD/RESET Position momentarily, then back to normal C is correct. EPP-9 step 3. Common Question 013 Tier 2 Group 1 KIA Importance Rating -RO 3.1 / SRO 3.17 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CSS controls including:

Containment humidity. Reference(s) -EPP-9 step 3 Proposed References to be provided to applicants during examination -None Learning Objective -Question Source -New Question History -Question Cognitive Level -Memory 10 CFR Part 55 Content -41 Comments -Category 1: Category 3: Category 5: Category 7: Tuesday, June 10,20081:11:02 PM Category 2: Category 4: Category 6: Category 8: 1 HLC-08 NRC Written Exam 55. Given the following: -A leak in the EDG "A" Air Start Receiver has reduced starting air pressure to 105 PSIG. -LCO 3.8.3, DIESEL FUEL AND STARTING AIR, Condition D, has been entered. -The leak has been repaired. As air pressure is restored, at which ONE (1) of the following pressures will the LCO be exited? A. 210 PSIG. B. 216 PSIG. C. 220 PSIG. D. 242 PSIG. 55 HLC-08 NRC Written Exam 55. 064 K6.07 OOllEMERG DIESEL GEN/2/1/2.7/2.9/ROILOW/N/NNEW -20081EDG-009 Given the following: -A leak in the EDG "A" Air Start Receiver has reduced starting air pressure to 105 PSIG. -LCO 3.B.3, DIESEL FUEL AND STARTING AIR, Condition D, has been entered. -The leak has been repaired. As air pressure is restored, at which ONE (1) of the following pressures will the LCO be exited? A'I 210 PSIG. B. 216 PSIG. C. 220 PSIG. D. 242 PSIG. The correct answer is A. A: Correct -LCO 3.B.3 is entered when pressure is 210 PSIG and 100 PSIG. B: Incorrect -216 PSIG is AIR START LO PRESS alarm setpoint. C: Incorrect -220 PSIG is the START pressure setpoint for Air Start Compressor in AUTO. D: Incorrect -242 PSIG is the STOP pressure setpoint for Air Start Compressor in AUTO. Exam Question Number: 55

Reference:

ITS 3.B.3, Condition D; ITS Bases 3. B .3, D1; SD-005, EDG, Page 27. KA Statement: Knowledge of the effect of a loss or malfunction of the following will have on the ED/G system: Air receivers. History: New -Written for HLC-OB NRC Exam. 62 --3.8 ELECTRICAL POWER SYSTEMS Diesel Fuel Oil, and Starting Air 3.8.3 3.8.3 Diesel Fuel Oil and Starting Air LCO 3.8.3 The common stored diesel fuel oil and starting air subsystem for each diesel generator (OG) shall be within limits. APPLICABILITY: When associated DG is required to be OPERABLE. ACTIONS ------------------------. ---------- .. NOTE ---------------------- .. ------------. Separate Condition entry is allowed for each OG. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more OGs with A.I Restore fuel oil 48 hours OG fuel oil level level to within < 19,000 gal and limits. > 14,145 gal in the Unit 2 DG fuel oil storage tank. B. One or more DGs with B.1 Restore fuel oil 48 hours DG Fuel oil level level to within < 34,000 gal and 1 imits. > 29,145 gal in the combination of the Unit 1 IC turbine fuel oil storage tanks and the Unit 2 DG fuel oil storage tank. (continued) HBRSEP Unit No. 2 3.8-16 Amendment No. 176 -ACTIONS (continued) CONDITION C. One or more DGs with new fuel oil not within imits. D. One or more DGs with starting air receiver pressure < 210 psig and == 100 psig. E. Required Action and associated Completion Time not met. OR -Common stored DGs diesel fuel oil or starting air subsystem for each DG not within limits for reasons other than Condition A. B. C. or D. -HBRSEP Unit No. 2 Diesel Fuel Oil. and Starting Air 3.8.3 REQUIRED ACTION COMPLETION TIME C.1 Restore stored fuel 30 days oil to within imits. .1 Rest9re starting air )48 hours recelver pressure to == 210 psig. E.1 Declare associated Immediately DG(s) inoperable. 3.8-17 Amendment No. 176 BASES ACTIONS SURVEILLANCE REQUIREMENTS HBRSEP Unit No. 2 C.l (continued) Diesel Fuel Oil and Starting Air B 3.8.3 may involve feed and bleed procedures. filtering. or combinations of these procedures. Even if a DG start and load was required during this time interval and the fuel oil properties were outside limits, there is a high likelihood that the DG would still be capable of performing its intended function. With starting air pressure < 210 psig, sufficient capacity for eight successive DG start attempts does not exist. However, as long as the receiver pressure is > 100 psig, there is adequate capacity for at least one start attempt, and the DG can be considered OPERABLE while the air recei ver is restored to the requ ired 1 i mi t . A period of 48 hours is considered sufficient to complete restoration to the required pressure prior to declaring the DG inoperable. This period is acceptable based on the remaining air start capacity, the fact that most DG starts are accomplished on the first attempt, and the low probability of an event during this brief period. E.1 With a Required Action and associated Completion Time not met, or one or more DG's fuel oil. or starting air subsystem not within limits for reasons other than addressed by Conditions A through D, the associated DGs may be incapable of performing its intended function and must be immediately declared inoperable. SR 3.8.3.1 This SR provides verification that there is an adequate inventory of fuel o'i 1 in the storage tanks to support one DG's operation for "7 days at full load. The 7 day period is sufficient time to place the unit in a safe shutdown condition and to br'ing in replenishment fuel from an offsite location. (continued) B 3.8-35 Revision No. 0 SD-005 EMERGENCY DIESEL GENERATOR SYSTEM , cylinder compressors (driven by A.C. motors), two air dryers (cooled by service water), two air storage reservoirs (34fe), pressure relief valves, check valves, shut-off valves, pressure gauges, and the necessary stop-start switches. The start-stop pressure .....::::> switches (PS-1961A(B>> are set to maintain220 reservoir pressure. The air compressor package may be started from MCC-5 (Diesel "A") and MCC-6 (Diesel liB "). The low air pressure alarm PS-4503A(B) (APLA) is set to give an alarm &11.2 b (rTf(, ",y' psig. A low air pressure alarm can be verified by the instrument panel gauge reading . . Engine starting is accomplished by supplying compressed air to 6 of the 12 cylinders in the correct firing order. Starting air is also supplied to two lube oil boosters (Governor and Upper Crank Bearings) when the engine is started. Technical Specification LCO 3.8.3 and SR 3.8.3.3 require starting air pressure to ;;:: 210 psig for EDG operability. 3.7.2 Air Start Valves and Air Start Solenoid Valves Two air start valves and air start solenoid valves are provided per engine. They are powered from MCC-A, Dist. Panel A (ckt 24) for EDG A, and MCC-B, Dist Panel B (ckt 12) for EDG B. They will close in 10 seconds or when> 200 rpm. Air can be cross tied by opening DA-28 (chain operated valve in "A" diesel room). A second isolation valve, DA-40, was installed in series with DA-28 to provide assurance of no train interaction between the A and B EDG starting air subsystems, downstream of the EDG starting air receivers. (EC52503 due to be completed 9/20104) 3.7.3 Hand Start Valves (DA-41, DA-42, DA-43, DA-44) There are two hand start valves on each emergency diesel generator. These valves can be used to start the diesel generator manually in case of electrical failure. The valves are normally open. The handle is a spring return to open handle. When a handle is moved to the closed positions, if held in that position, it will vent the air from the outside of the elastomeric tube of the associated DG air start valve, causing the air start valve to open and admitting starting air to the EDG cylinders. The valves can also be used for rotating the engines without starting them if proper precautions are taken. 3.7.4 Air Start Control Valve This valve has no internals and simply functions as a piece of pipe (per discussion with Fairbanks Morse). 3.7.5 Air Start Distributor EDG There is one cam operated distributor per engine which provides pilot air to open the air start check valves for each supplied cylinder. As cam rotates, distributor valves are opened against spring pressure, according to firing order. Page 27 of 74 Revision 11 INFORMATION USE ONLY HLC-08 NRC Written Exam 56. Given the following: -The plant has been restarted following a forced outage. -Loop flow measurements have determined that RCP "B" impeller has degraded such that its RCS loop flow has DECREASED by 5% from its original value. -The other RCS loop flows remain UNCHANGED. Which ONE (1) of the following would be a result of the decreased flow rate in Loop "B"? A. Demand on the PZR Control Group heaters at 2235 PSIG will be lower. B. The reactor core will operate closer to DNB when at full power. C. Core Delta-T at full power will be lower. D. The reactor core will operate further from DNB when at full power. 56 HLC-08 NRC Written Exam 56. 002 K6.07 001lREAC COOL PUMP/2/2/2.512.8IROILOW/N/A/BYRON-2003/THERMO CHAP 8-014 Given the following: -The plant has been restarted following a forced outage. -Loop flow measurements have determined that RCP "s" impeller has degraded such that its RCS loop flow has DECREASED by 5% from its original value. -The other RCS loop flows remain UNCHANGED. Which ONE (1) of the following would be a result of the decreased flow rate in Loop "S"? A. Demand on the PZR Control Group heaters at 2235 PSIG will be lower. B:t The reactor core will operate closer to DNS when at full power. C. Core Delta-T at full power will be lower. D. The reactor core will operate further from DNS when at full power. The correct answer is S. A: Incorrect -RCPs "s" and "C" supply PZR spray valves. However, based on plant design, RCP "s" provides limited flow, plus under steady state/2235 PSIG conditions, the spray valves are closed and bypasses provide minimal flow to keep the lines warm -NOT an impact on current draw on the PZR Control Group heaters. S: Correct -Putting out the same MWt with a reduced flow rate means reduced heat transfer capabilities and therefore operation closer to DNS. C: Incorrect -Delta-T should actually be higher in this situation. D: Incorrect -Mass flow rate reduction for the same power results in operation CLOSER to DNS. Exam Question Number: 56

Reference:

Thermodynamics, Chapter 8, Pages 40-41. KA Statement: Knowledge of the effect or a loss or malfunction on the following RCS components: Pumps. History: Direct from Sank. 63 ----,. -....... --I INSTRUCTOR GUIDE I d. As steam bubble is swept from fuel : \ rod surface it agitates laminar layer, increasing turbulent flow e. The mixing of coolant in turbulent flow increases heat transfer rate f. Reactor coolant flowrate affects heat transfer rate in fuel channel g. When heat is transferred from fuel cladding to coolant, it must pass through laminar flow region before reaching turbulent bulk flow region h. The thickness of laminar region is proportional to its resistance to heat flow i. As flow rate increases, outer portions of laminar layer are disturbed, resulting in more turbulent flow, thinner laminar layer and increased heat transfer coefficient ! , i I KEY POINTS, AIDS, QUESTIONS/ANSWERS 1 3. Reactor coolant flowrate also has direct affect on fuel temperature f , Objective 18.e 4. a. Rate of heat transfer is proportional to reactor coolant flowrate and differential temperature between fuel and coolant: Where: a. Q = heat transfer rate (Btu/hr) b. ri1 = mass flow rate (lbn/hr) c. c p = specific heat capacity for coolant at constant pressure (B tu/lb m OF) d. T = differential temperature between fuel and coolant (OF) , ; PWR / THERMODYNAMICS / CHAPTER 8 / THERMAL HYDRAULICS 40 of 69 Equation 8-7 I TP 8-29 Q = © 1999 GENERAL PHYSICS CORPORATION REV 3 INSTRUCTOR GUIDE; .. 5. If reactor power is held constant, reducing reactor coolant flow rate increases differential temperature

a. Fuel temperature increases to make ; up for decreased reactor coolant ' flowrate 6. Heat flux in fuel rod is proportional to neutron flux present in that region a. The heat flux for any point in fuel is also proportional to neutron flux at that point b. Thus heat flux, neutron flux, and heat generation rate are proportional in fuel rod 7. This is shown in Figure 8-16, which shows shapes of neutron flux and heat generation from centerline to edge of fuel pellet a. Note that no heat generation is shown in clad and gap, and that small amount of heat is generated in coolant by gamma and neutron heating ; 8. Fuel densification is another factor that i can affect overall L\ T and thus centerline temperature

a. During initial stages of nuclear fuel operation, fuel becomes denser, increasing gap between fuel pellets and cladding b. This can change L\T across gap, and thus centerline temperature:

9. Where: PWR / THERMODYNAMICS

/ CHAPTER 8 / THERMAL HYDRAULICS 41 of 69 KEY POINTS, AIDS, QUESTIONS/ANSWERS Figure 8-16/ TP 8-30 COOLANT (WATER) /lfo<t,-'--ZIRCAI.OY -4 CLADDING , HELlUM*PRESSURIZED GAP Equation 8-8/ TP 8-31 Q rug =-x-gap A k g g © 1999 GENERAL PHYSICS CORPORATION REV 3 Quest No: 1 ROSRO: Both TIER: 2 GROUP: 1 Topic No: 003000 "Bv rot1 t 0 :5 KANo: 003000KS.0l Category Statement: RO: 3.3 SRO: 3.9 Cog Level: High SystemlEvolution Name: Reactor Coolant Pump System (RCPS) Knowledge of the operational implications of the following concepts as they apply to the RCPS: KA Statement: The relationship between the RCPS flow rate and the nuclear reactor core operating parameters (quadrant power tilt, imbalance, DNB rate, local power density, difference in loop T-hot pressure) UserID: BY03NRC-OOI Topic Line: -The plant was restarted following a forced outage. -The plant was restarted following a forced outage. Question Stem: _ Loop flow measurement has determined the 'c' RCP impeller has degraded such that its RCS loop flow has DECREASED by 5% from its original value. -The other three RCS loop flows remain UNCHANGED. Which one of the following would be a result of the decreased flow rate in the 'c' loop? Demand on the Pressurizer variable heaters at 2235 psig will be lower. A: The reactor core will operate closer to DNB when at full power. B: Delta-Tin the 'c' RCS loop at full power will be lower. C: Steam pressure in the 'c' SIG at full power will be higher. D: Answer: Task No: B Obj No: A.PD3-07 Time: Cross Ref: Question Source: modified: {from Kewaunee 12118/97, question 10 10835 in INPO eaxm bank -modifed stem to apply to 4 loop plant and replaced distractor A.} SD reviewed format OK 7/18103, tech review OK by TL 7/22103

Reference:

ILT Fundamentals Lesson Plan Power Distribution Chapter 3, Thermal Limits; page 15-16. Il-PD-XL-03 A: INCORRECT- 'c' & 'D' RCPs supply pzr spray. However, based on plant design, the 'c' RCP provides limited flow, plus under Explanation: Question Difficulty Medium steady statel2235 psig conditions, the sprays are closed and bypasses provide minimal flow to keep the lines warm -not an impact on current draw on the variable heaters. B: CORRECT -putting out the same MWt with a reduced flowrate means reduced heat tranfer capabilities and therefore operation closer toDNB. C: INCORRECT -Delta-T should actually be higher in this situation. D: INCORRECT -Higher Delta-T means Tcold should be lower and therefore steam pressure should be lower. Date Written: 7/14/2003 Author: T. Foss App. Ref: None HLC-08 NRC Written Exam 57. During withdrawal of the Control Banks, APP-005-F2, ROD BOTTOM / ROD DROP will CLEAR when Control Bank "A" reaches ... A. 20 steps as sensed by the Pulse To Analog Converter. B. 35 steps as sensed by the Pulse To Analog Converter. C. 20 steps as sensed by the IRPls. D. 35 steps as sensed by the IRPls. 57 HLC-08 NRC Written Exam 57.014 K1.01 OOlfROD POSITION INDICAT/2/2/3.2/3.6fRO/LOWININNEW -200S/IRPI-007 During withdrawal of the Control Banks, APP-005-F2, ROD BOTTOM / ROD DROP will CLEAR when Control Bank "A" reaches ... A. 20 steps as sensed by the Pulse To Analog Converter. B. 35 steps as sensed by the Pulse To Analog Converter. 20 steps as sensed by the IRPls. D. 35 steps as sensed by the IRPls. The correct answer is C. A: Incorrect -The P to A converter only inputs to the Rod Bottom Bypass Bistable. Control Bank A does NOT have Rod Bottom Bypass Bistable. B: Incorrect -The P to A converter only inputs to the Rod Bottom Bypass Bistable. Control Bank A does NOT have Rod Bottom Bypass Bistable. Setpoint of 35 steps is for Bypass Bistable setpoint. C: Correct -Control Bank "A" does NOT have a Rod Bottom Bypass Bistable. APP-005-F2 will clear when ALL Control Bank "A" RBLs clear at 20 steps. 0: Incorrect -Control Bank A does NOT have Rod Bottom Bypass Bistable. 35 steps is the <__ Rod Bottom Bypass Bistable setpoint. Exam Question Number: 57

Reference:

SD-009, IRPI, Pages 16, 17 and Figure 2; APP-005-F2. KA Statement: Knowledge of the physical connections and/or cause-effect relationships between the RPIS and the following systems: CRDS. History: New -Written for HLC-08 NRC exam. 64 SD-009 INDIVIDUAL ROD POSTION INDICATION SYSTEM 4. Rod Bottom Lamp Outputs 1. Rod Bottom Relay 2. Meter 3. Computer 4. Annunciator 24 VAC (Nominal) Capable of driving a 12 VAC coil (Struthers Dunn PN#A314XCX48P-12A) Capable of driving a panel meter with impedance >1.5K ohms +0.075 to -3.67 VDC Input Capable of driving a computer input module with impedance> lOOK ohms +0.075 to -3.67 VDC Contact closure Switch rating:

• Low level/dry circuit or power 0.4 VA Max @ 20 vac or DC Max or 5 amps @ 120 V AC or 28 VDC
• Recommended loading: 0.4 VA 3.2.2.3 Environment Ambient Temperature Normal operating temperature Abnormal operating temperature Ambient Humidity Normal operating range Abnormal operating range <12 Hrs. Accuracy 85°F +/- 20°F 50°F to 140° 60% (Nominal) o to 95 % non-condensing

+/-3 Steps or +/-O. 045 .VDC 3.2.3 P-A Bypass, Micro Sentry Modules (Control Bank Bypass Bistable) (Refer to Figure-2 & 19) The three modules for the Control Bank Bypass Bistables are located in Rack 2 above the P-A converter. There is a bypass bistable associated with Control Bank B, C, and D. The Shutdown Banks and Control Bank A does not have bypass bistables. The bypass bistable obtains its signal from the analog signal out of the P-A Converter corresponding to the demand position for the banks. The bypass bistable was originally designed to provide a blocking action on the turbine IRPI Page 16 of 27 Revision 2 INFORMATION USE ONLY SD-009 INDIVIDUAL ROD POSTION INDICATION SYSTEM runback signal, rod bottom/rod drop annunciator, and the inhibiting of the automatic rod withdrawal when the rods of the associated bank are to be operated near or below the Rod Bottom Bistable setpoints. Bypass of these functions were considered necessary in the original design of the plant to allow automatic withdrawal of the rods during plant startup and to prevent undesired runbacks when operating near or below the Rod Bottom Bistable setpoints. If the Rod Bottom Bistables were not bypassed, the rod drop contact in the in-out relay circuits of the Rod Control System would prevent automatic rod withdrawal since any banks not yet moved would indicate a rod bottom. A bypass was not considered necessary for the Shutdown Banks or Control Bank A since these banks was determined to always be withdrawn prior to placing the control system in automatic. Automatic Rod withdrawal has been permanently disable and the turbine runback has been eliminated. This blocking feature still exists in the RPI System but it's ONLY function now is to block the rod bottom/ if conditions are met. Note: If control rods are being withdrawn in Control Bank B,C, or D and the demanded position of the bank :::;; 35 steps and a rod drops in the bank, The RTGB rod drop annunciator would NOT alert the Operator. However, the rod bottom LED would still illuminate. The bypass bistable is set to operate at 35 steps from the bottom of the core. The blocking action is automatically removed when the control bank demand position is above 35 steps (+ 3 steps for reset). These bypass bistables would also operate if the Pulse to Analog Converter lost power or had an input < 35 steps demanded rod position.(example: rod control startup pushbutton inadvertently pushed at power) 3.2.4 Pulse to Analog (P-A) Converter Drawer (RPI-Figure-20 & 21) IRPI The P-A Converter Drawer is located on a single chassis in RPI Cabinet #2. This single drawer actually contain four (4) P-A Converters, one for each Control Rod Bank. The P-A Converter receives up or down pulses from the slave cycler in the Rod Control System. P-A Converter converts these pulses to a DC analog signal. This signal is sent to Rod Bottom Bypass Bistable, to rod insertion limit circuit (alarms), and to the local indicator. In addition to detecting the actual vertical position of each CRDM drive rod with the IRPI System, the desired position of the control rods based on command signals from the Rod Control System is required. The demand signals from the Rod Control System drive mechanical step counters on the main control board and provide inputs to the plant computer and the P-A Converter. The function of the P-A Converter is to count Control Bank A, B, C and D step pulses for the Rod Control System and convert these step counts into an equivalent analog DC voltage. The DC voltage is then sent to the rod insertion limit (RIL) computer in the process instrumentation, to the bank D Page 17 of 27 Revision 2 INFORMATION USE ONLY TUT SIGNAL ) Ip Vsec ROD CONTROL sTAlttUP SIlTCH IUSTIR CYClJ:R SLAVE CYCLER SIGNAL CONJ)mOHlNC IroDUU: 1-RPICHANNEL RPI-FIGURE-2 RTGB INDICATOR 'osmON SICHAL ... _ 'l'O PLAJI'f -COMPUTIR 'i<IU.--II UrI:I"AU::/-'i<AI..-" ... "RJ".t. IN TES1' -----.. -... coJrraOLROOIi MEDIAN &' FROM ERFIS PO$ITlON SIG:NA1. SHon IP.A on Control Banks ONLY IN ROD BOTfO)( COHJ)moN JU.HUAL SE'I1'ER. ...----... PULSE TO ANALOe COtNATER 1'0 R.r.L CU. (ALARKS) toc.V. INDICATION TO COMPUTIR INFORMATION USE ONLY -S1'CP COUM'tIR ALARM ROD BOTTOM ROD DROP AUTOMATIC ACTIONS 1. None CAUSE 1. Control Rod dropped (IRPI) 2. IRPI Failure OBSERVATIONS 1 . Rod Bottom Light on Dropped Rod 2. APP-005-A3, PR DROPPED ROD annunciator

3. Power Range meters and Recorder ACTIONS 1. IF a dropped rod OR an IRPI failure has occurred, THEN refer To AOP-001. DEVICE/SETPOINTS

.-...::::;;:.-

1. Individual RPI Rod Bottom Switches /20 steps from fully inserted POSSIBLE PLANT EFFECTS 1. Power Reduction

2. Flux Tilt 3. Entry into TECH SPEC LCO Action REFERENCES

1. ITS LCO 3.1.7 and ITS LCO 3.1.4 2. AOP-001, Malfunction of Reactor Control System 3. CWO B-190628, Sheet 80 Cable AS I APP-005 Rev. 29 APP-005-F2 Page 35 of 40 I HLC-08 NRC Written Exam 58. Given the following: -A Reactor Startup is in progress. -I R Channel N-35 indicates 3E-1 0 amps. -IR Channel N-36 indicates 9E-11 amps. -No manual blocks have been inserted.

-SR Channel N-32 indicates 8100 cps. -SR Channel N-31 fails HIGH. Which ONE (1) of the following describes the plant response to the SR Channel N-31 failing HIGH? A. A reactor trip signal is generated, resulting in a reactor trip. B. A reactor trip signal is generated, but no trip occurs since ONE (1) IR channel is above P-6. C. No reactor trip signal is generated since ONE (1) IR channel is above P-6. D. No reactor trip signal is generated, but the N-31 Level Trip Switch must be taken to BYPASS. 58 HLC-08 NRC Written Exam 58. 015 K4.01 001lNUCLEAR INSTRUMENT/2/2/3.1I3.3IROIHIGHINIA/FARLEY -2001INIS-01O Given the following: -A Reactor Startup is in progress. -IR Channel N-35 indicates 3E-10 amps. -IR Channel N-36 indicates 9E-11 amps. -No manual blocks have been inserted. -SR Channel N-32 indicates 8100 cps. -SR Channel N-31 fails HIGH. Which ONE (1) of the following describes the plant response to the SR Channel N-31 failing HIGH? A'I A reactor trip signal is generated, resulting in a reactor trip. B. A reactor trip signal is generated, but no trip occurs since ONE (1) IR channel is above P-6. C. No reactor trip signal is generated since ONE (1) IR channel is above P-6. D. No reactor trip signal is generated, but the N-31 Level Trip Switch must be taken to BYPASS. The correct answer is A. A: Correct -Since no blocks are in for the SR, a Reactor Trip will occur. B: Incorrect -The trip must be manually blocked when above P-6. c: Incorrect -Although above P-6, the SR trip signals have NOT been manually blocked. 0: Incorrect -The trip must be manually blocked when above P-6, taking N-31 Level Trip Switch to BYPASS is the correct action per the OWP, but the Reactor will trip. Exam Question Number: 58

Reference:

SD-011, RPS, Figure 18. KA Statement: Knowledge of NIS design feature(s) and/or interlock(s) provide for the following: Source-Range detector power shutoff at high powers. History: Direct from Bank. 65 SOURCE RANGE HIGH FLUX REACTOR TRIP RPS-FIGURE-18 SOURCE RANGE HIGH FLUX REACTOR TRIP CHANNELl TRIP BYPASS (NISRACK) NOT ENERGIZE SOURCE RANGE H.V. (EITltER LOGIC TRAIN) CHANNEL II MANUAL BLOCK (MOMENTARY) l DEENERGIZESOURCE RANGE H.V. (80TH A&B LOGIC TRAIN REQUIRED) NOT REACTOR TRIP TRIP BYPASS (NISRACK) NOT I rpsf18 I B. 79%. C.73%. D.70% For every 1 % above QPTR of 1.00, reduce power by 3%, when QPTR exceeds 1.02. A -Incorrect, QPTR UPPER: 1.0176; 1.0176-1.00=1.76% (rounded to 2%); 100% (3%*2% )=94%. B -Incorrect, 1.07-1.00=7%; 100%RTP-(3%*7%)=79%. C -Correct, QPTR LOWER: Average of lower excore detectors is 0.98. 1.07/0.98=1.0918; 1.0918-1.00=9%; 100%RTP-(3%*9%)=73%. D -Incorrect, 1.09-1.00=9%; 97%RTP-(3%*9%)=70% Source: Farley NRC Exam 2000-301 Answer: C 33. 015A2.011 l-tLY'Le,y ':;2. 00 I Given the following conditions: -Reactor startup is in progress. -All NI switches are in their normal lineup. -Intermediate Range (IR) Channel N-35 indicates 3E-1O. -IR Channel N-36 indicates 9E-11. -No manual blocks have been inserted. -Source Range (SR) Channel N-31 has failed high. -SR channel N-32 is functioning normally. Which ONE of the following describes the plant response to the SR Channel N-31 failing high? A. A reactor trip signal is generated resulting in a reactor trip. B. A reactor trip signal is generated, but no trip occurs since one IR channel is above P-6. c. No reactor trip signal is generated since one IR channel is above P-6. D. No reactor trip signal is generated but, the level trip switch must be taken to bypass as soon as N-36 indicates above 1E-1O. A -Correct, Since no blocks are in for the SR a reactor trip occurs. B -Incorrect, The trip must be manually blocked when above P-6. C -Incorrect, Although above P-6 the SR trip signals have not been blocked. D -Incorrect, The trip must be manually blocked when above P-6. HLC-08 NRC Written Exam 59. Given the following: -TI-579, Inadequate Core Cooling Monitor (ICCM) Train A Plasma Display, is deenergized. -A Loss of Offsite Power and Reactor Trip occur. -Natural Circulation conditions are being verified in Supplement E, NATURAL CIRCULATION VERIFICATION. How will Subcooling and Core Exit Thermocouple Temperatures (CETC) be determined? A. Subcooling from the Train A Subcooling Monitor. CETC temperatures by local readings on the junction boxes. B. Subcooling by comparing highest Hot Leg temperature to highest RCS wide range pressure. CETC temperatures by thermocouple readings. C. Subcooling by ICCM Train B. CETC temperatures by BOTH Train A and Train B CETCs. D. Subcooling by comparing Hot Leg temperatures to Cold Leg temperatures. CETC temperatures by Train B CETCs ONLY. 59 HLC-08 NRC Written Exam 59.017 K3.01 001/IN-CORE TEMP MONITORl2/2/3.5/3.7IROIHIGH/NINNEW -20081ICCM-005 Given the following: -TI-579, Inadequate Core Cooling Monitor (ICCM) Train A Plasma Display, is deenergized. -A Loss of Offsite Power and Reactor Trip occur. -Natural Circulation conditions are being verified in Supplement E, NATURAL CIRCULATION VERIFICATION. How will Subcooling and Core Exit Thermocouple Temperatures (CETC) be determined? A. Subcooling from the Train A Subcooling Monitor. CETC temperatures by local readings on the junction boxes. B. Subcooling by comparing highest Hot Leg temperature to highest RCS wide range pressure. CETC temperatures by thermocouple readings. Subcooling by ICCM Train B. CETC temperatures by BOTH Train A and Train B CETCs. D. Subcooling by comparing Hot Leg temperatures to Cold Leg temperatures. CETC temperatures by Train B CETCs ONLY. The correct answer is C. A: Incorrect -CETCs do NOT have any local indication at the junction boxes. B: Incorrect -Subcooling should be determined from the highest T HOT and lowest pressure. C: Correct -CETC can be displayed on CET Recorder and ERFIS if Plasma Display panel is inoperable. D: Incorrect -Hot leg temperature to cold leg temperature equals Delta T. Exam Question Number: 59

Reference:

Supplement E, Page 29; ICCM Figure 1, 27. KA Statement: Knowledge of the effect that a loss or malfunction of the ITM system will have on the following: Natural circulation indications. History: New -Written for HLC-08 NRC Exam. 66 EPP-Supplements INSTRUCTIONS SUPPLEMENTS CONTINUOUS USE Supplement E Rev. Page RESPONSE NOT OBTAINED Natural Circulation Verification (Page 1 of 1) 1. Check Natural Circulation Status Increase dumping steam. As Follows:

• Steam pressure -STABLE OR DECREASING
• RCS subcooling

-GREATER THAN 35°F [55°F]

• Core exit TiCs -STABLE OR DECREASING
• RCS hot leg temperatures

-STABLE OR DECREASING

• RCS cold leg temperatures

-TRENDING TO SATURATION TEMPERATURE FOR STEAM PRESSURE 2. Return To Procedure And Step In Effect -END -35 29 of 89 RVLIS TRAIN AI TRAIN B I CETM TRAIN AI TRAIN B 1 SMM TRAIN A TRAIN B ICCM SIMPLIFIED TRAINS ICCM-FIGURE-l (Rev. 1) ICCM r---tI TRAIN A ICCM PLASMA ELECTRONICS DISPLAY CABINET * * ------ TRAIN A Control Room BAY -II ---r-l' TRAIN B BAY control Room Auxilary Building Rod Drive Control Room I ICCM TRAIN B L.I PLASMA DISPLAY INFORMATION USE ONLY ICCM-86 SIGNAL PROCESSOR INPUTS/OUTPUTS ICCM-FIGURE-27(Rev.2) INPUTS 2 WIDE RANGE TEMP. (HOTS) WIDE RANGE PRESS. 2 TOP OF VESSEL RTD HOT LEG RTD 2 BOTTOM .OF VESSEL RTD TOP OF VESSEL HYDRAULIC ISOLATOR HOT LEG HYDRAULIC ISOLATOR BOTTOM OF VESSEL HYDRAULIC ISOLATOR UPPER RANGE DIP TRANS. FULL RANGE DIP TRANS. DYNAMIC HEAD DIP TRANSMITTER 1 WIDE RANGE TEMP. (HOT) WIDE RANGE PRESS. 2 TOP OF VESSEL RTD BOTTOM .OF VESSEL RTD TOP OF VESSEL HYDRAULIC ISOLATOR HOT LEG HYDRAULIC ISOLATOR BOTTOM OF VESSEL HYDRAULIC ISOLATOR UPPER RANGE DIP TRANS. FULL RANGE DIP TRANS. DYNAMIC HEAD DIP TRANSMITTER TE-413-2 TE-423 .. p PT-511AA ... TE-511AA po TE-511AB ... po TE-511AC .. ... TE-511AF ICCM-86 TE-511AG ... CHANNEL I po LlS-511AB .. ... LlS-511AA .. p LlS-511AC .. p LT-511AA _ .. LT-511AB po ... ... LT-511AC .. TE-433 ... po PT-511BA ... TE-511BA po TE-511BB ... po ICCM-86 CHANNEL II TE-511 BF .. p LlS-511BB .. ... LlS-511BA ... p LlS-511 BC ... LT-511 BA ... LT-511 BB .. ... LT-511BC .. po INFORMATION USE ONLY OUTPUTS PLASMA DISPLAY PANEL .. ... 8 CET RECOR2,.ER ... DATA LINK TO ERFIS ... p PLASMA DISPLAY PANEL .. ... 8 CET RECORD,R DATA LINK TO ERFIS ... I ICCMF27 I HLC-08 NRC Written Exam 60. Given the following: -An RCS LOCA has occurred which resulted in a Containment Spray Actuation. -ALL ESF equipment has functioned as designed. -NO ESF signals have been reset. -Containment pressure is 8 PSIG. The REQUIRED action(s) the crew must perform in order to allow valve SI-845A, SAT DISCH VLV, to be CLOSED, is(are) to reset the ... A. SI Signal, and then reset the Containment Spray Actuation Signal. B. Containment Spray Actuation Signal. C. Containment Isolation Phase B Signal. D. SI Signal, and then reset the Containment Isolation Phase B Signal. 60 HLC-08 NRC Written Exam 60. 027 Kl.Ol OOl/CTMT IODINE REMOV AL/2/2/3.4/3.7IROIHIGH/NINBRAIDWOOD -20041ESF-004 Given the following: -An RCS LOCA has occurred which resulted in a Containment Spray Actuation. -ALL ESF equipment has functioned as designed. -NO ESF signals have been reset. -Containment pressure is 8 PSIG. The REQUIRED action(s) the crew must perform in order to allow valve SI-845A, SAT DISCH VLV, to be CLOSED, is(are) to reset the ... A. SI Signal, and then reset the Containment Spray Actuation Signal. B:I Containment Spray Actuation Signal. C. Containment Isolation Phase B Signal. D. SI Signal, and then reset the Containment Isolation Phase B Signal. The correct answer is B. A: Incorrect -SI reset is unnecessary, therefore it is NOT a required action. B: Correct -CV Spray Actuation can be reset since CV pressure is less than 10 PSIG. C: Incorrect -Resetting Phase B would have NO effect on CV Spray. D: Incorrect -Resetting SI and Phase B would have NO effect on CV Spray. Exam Question Number: 60

Reference:

SD-006, ESFAS, Pages 13,17 and Figure 3. KA Statement: Knowledge of the physical connections and/or cause-effect relationships between the CIRS and the following systems: CSS. History: Direct from Bank. 67 SD-024 CONTAINMENT SPRAY SYSTEM 3.3 Eductors Number Volume Design Pressure Design Temperature Flow (max) 2 (1 for each pump) Provide enough NaOH to keep pH.2:.. 9.3 200 psig Ambient 80gpm The purpose of the eductors is to provide a low pressure area for injection of the NaOH solution stored in the SAT. The eductors use the venturi principle for injecting NaOH into the suction of the spray pumps. As the velocity increases the pressure decreases at the constricted section. The SAT is connected at this low pressure area. 3.4 Major Valves 3.4.1 Suction Relief Valve, SI -871 The purpose of the SI-871 is to protect the suction piping from overpressurization. The relief valve is set at 200 psig. It relieves to "B" Spray Header downstream of SI-880C and SI-880D. 3.4.2 CV Spray Pump Suction Valves, SI-844A and SI-844B The purpose of the suction valves is to allow pump isolation for maintenance. They are motor operated and controlled with a close/open switch from the RTGB. At power, these valves are normally open. SI-844A and SI-844B are powered from MCC-5 and MCC-6, respectively. Spray Additive Tank Discharge Valves, SI-845A and SI-845B The purpose of the SAT discharge valves is to isolate the SAT from the eductors. These parallel valves are motor operated and are normally closed, but will open automatically on spray signal (P-signal). They are controlled from the RTGB with a Close/ Auto/Open switch (spring return to Auto from open) and are powered from MCC-5 for SI-845A and MCC-6 for SI-845B. 3.4.4 Spray Additive Tank Throttling Valve, SI-845C css The Spray Additive Tank throttling valve is used to adjust the flowrate of NaOH to the eductor thereby controlling the concentration of NaOH in the spray being delivered to the CV. It is a normally open, motor operated valve powered from MCC-5 and is controlled with a close/open switch (spring return to center) located on the RTGB. Page 13 of26 Revision 9 INFORMATION USE ONLY SD-024 Reset of Containment Spray A Containment Spray signal can be reset (key switch) after actuation if it becomes necessary to stop or realign equipment actuated by the Containment Spray signal. Resetting Containment Spray will not "terminate" Containment Spray. The components actuated by Containment Spray do not change state when Containment is reseL-J'hey must be manually aligned after Containment Spray is reset. During thlS"Condition, if another Containment Spray signal is generated, a manual Containment Spray must be initiated by the operator for Containment Spray to occur. Until the Containment Spray signal is manually reset, any safeguards equipment stopped from the RTGB cannot be restarted without removing the control power fuses at the breaker and reinstalling them. This is due to the anti-pump feature in the breakers. The Phase "B" Containment Isolation must be reset individually after the Containment Spray signal is reset or cleared. Phase "B" Containment Isolation is reset from pushbuttons on the RTGB. Manual spray actuation will cause the following:

1) Spray actuation

2) Containment Phase "B" -3) Containment Ventilation Isolation

-The following valves will close:

• Purge Valves
• Pressure Relief Valves
• Vacuum Relief Valves 5.2 Auto-actions on Spray Actuation

1) 2)

Both Spray Pumps start Spray Pump Discharge Valves (880 A/B/C/D) Open Spray Additive Tank Outlets (845 AlB) Open 6.0 SYSTEM OPERATION 6.1 Normal Operation css The Spray system is in standby whenever plant is above cold shutdown conditions. The pump discharge and spray tank outlet valves are shut. The Spray pumps are aligned for auto start with the suction valves opened to the RWST. There is a note in PLP-111, Leakage Reduction Program, for plant personnel to notify the Control Room for eight different systems when leaks are identified on components Page 17 of26 Revision 9 INFORMATION USE ONLY Lege.nd: @ Pressure PC Channel 951A U .ORGate ..w,.. Coincidence (2 out of 3) TI AND Gate

• CD Retentive Memory LOGIC DIAGRAM ON SPRAY ACTUATION CSS-FIGURE-3 2/3 I Hi-Hi Steam tine /sl SIGNAL
• Isolation /
• M.R.
• I 2/3 IHi-Hi INFORMATION USE ONLY Two Push Button Manuaf Spray Actucation Cont.lsol, Phcise 8 Cont. Veni Isol. I . Containment V"ntilatiori Isolation CSSF03 Quest No: RO SRO: TIER: GROUP: Topic No: KA No: RO: SRO: Cog Level: 59 RO 2 2 027000 0270ooK1.01 3.4 3.7 !,.ow System/Evolution Name: Category Statement:

Containment Iodine Removal System (CIRS) Kuowledge of the physical counections and/or cause-effect relationships between the CIRS and the followiug systems: KA Statemeut: CSS UserID: BW04NRC-059 Topic Liue: Question Stem: Given: -A Unit 1 RCS LOCA and Containment Spray Actuation have occurred. -ALL U1 ESF equipment has functioned as designed. -NO U1 ESF signals have been reset. The MINIMUM action(s) the crew must perform in order to allow 1 CS019A, EDUC 1 A SPRAY ADD VLV to CLOSE and remain CLOSED, is/are to reset the U1 Train A: A SI Signal, and then reset the U1 Train A Containment Spray Actuation Signal. B Containment Spray Actuation Signal, only. c Containment Isolation Phase B Signal, only. o SI Signal, and then the U1 Train A Containment Isolation Phase B Signal. Auswer: Task No: R-EP-009 Questiou Source: B Obj No: S.CSI-08-D Time: 1

Reference:

Cross Ref: BWLI-CSI-040 Big Note CS-I Rev 7 Explanation: NEW. Added unit references and the word "minimum" to the stem make sure only one correct answer. Got rid of Phase A as a choice, added SI signal. Question Difficulty 3 Containment spray actuation signal must be reset (nothing else). Phase B, and SI status have uo effect, though both are plausible since they both should have occurred. A: SI reset is unnecessary. B: Correct C: Phase B would have no effect. D: SI and Phase B resets would have no effect Date Written: 3/2212004 Author: Unknown App. Ref: None. HLC-08 NRC Written Exam 61. Given the following: -The plant is in MODE 6. -Refueling activities are in progress. -APP-005-C1, SR HI FLUX AT SHUTDOWN is illuminated. Which ONE (1) of the following describes SR counts when the alarm was received and the automatic action expected? SR counts are greater than ... A. 3 times AII-Rods-In background and the CV Evacuation horn will sound. B. 3 times AII-Rods-In background and the LOCAL Evacuation alarm will sound. C. 2 times AII-Rods-In background and the CV Evacuation horn will sound. D. 2 times AII-Rods-In background and the LOCAL Evacuation alarm will sound. 61 HLC-08 NRC Written Exam 61. 034 A4.02 OOllFUEL HANDLING EQUIP/2/2/3.5/3.9fROILOWINIAfNEW -200SINIS-00S Given the following: -The plant is in MODE 6. -Refueling activities are in progress. -APP-005-C1, SR HI FLUX AT SHUTDOWN is illuminated. Which ONE (1) of the following describes SR counts when the alarm was received and the automatic action expected? SR counts are greater than ... A'I 3 times AII-Rods-In background and the CV Evacuation horn will sound. B. 3 times AII-Rods-In background and the LOCAL Evacuation alarm will sound. C. 2 times AII-Rods-In background and the CV Evacuation horn will sound. D. 2 times AII-Rods-In background and the LOCAL Evacuation alarm will sound. The correct answer is A. A: Correct -High Flux at Shutdown alarm setpoint is set at 3 times the ALL-Rods-In background count rate to ensure early warning of changing counts. IF the setpoint is exceeded, an automatic initiation of the CV Evacuation Horn will occur. B: Incorrect -High Flux at Shutdown alarm setpoint is set at 3 times the ALL-Rods-In background count rate to ensure early warning of changing counts. There is NO automatic initiation of LOCAL evacuation. C: Incorrect -If SR counts double during any reactivity addition, dilution or rod movement is stopped to ensure positive control over the core. This is an incorrect application of this setpoint. D: Incorrect -If SR counts double during any reactivity addition, dilution or rod movement is stopped to ensure positive control over the core. This is an incorrect application of this setpoint. Exam Question Number: 61

Reference:

G P-002, Page 118; OP-002, Pages 15 and 16; APP-005-C1. KA Statement: Ability to manually operate and/or monitor in the control room: Neutron levels. History: New -Written for HLC-08 NRC Exam. 68 ALARM SR HI FLUX AT SHUTDOWN AUTOMATIC ACTIONS 1 . CV Evacuation Alarm CAUSE 1. Addition of positive reactivity while shutdown 2. Failure to block prior to startup 3. Electrically induced "noise" in SR Channel 4. Failure Source Range Channel OBSERVATIONS

1. Source Range NI ACTIONS 1. IF a valid High Flux level is indicated, THEN perform the following:

a. Evacuate Containment.

b. Insert any Control Rods withdrawn.

c. IF fuel movement is in progress, THEN discontinue fuel movement.

APP-005-C1

d. IF Source Range count rate is increasing, THEN borate the RCS lAW OP-301 until SUR becomes negative.

2. IF due to failure to block the alarm during startup, THEN block alarm. 3. IF due to electrical spikes AND personnel are in the CV, THEN make PA announcement to disregard alarm. 4. IF NOT due to known core reactivity change, THEN notify Reactor Engineering to evaluate the potential need to perform EST-001. (ACR 93-00198)

DEVICE/SETPOINTS

1. N-31, or N-32 I 3 times Countrate with all rods inserted. (CR 95-00294)

POSSIBLE PLANT EFFECTS 1. Inadvertent Criticality

2. Loss of required Technical Specification Shutdown Margin REFERENCES

1. ITS LCO 3.1.1 2. EST-001, Source Range Statistical Reliability Test 3. ACR 93-00198, N-32 failed to indicate proper change in countrate

4. CWD B-190628, Sh 441, Cable AM 5. CR 95-00294, IN 93-32: Nonconservative Inputs For Boron Dilution Event 6. OP-301, Chemical and Volume Control System (CVCS) ! APP-005 Rev. 29 Page 16 of 40 I 8.4.1.2 (Continued)

Section 8.4.1 Page 2 of 5 INIT INIT N-31 N-32 NOTE: Unless otherwise stated, the switches, potentiometers, AND indicators addressed by the procedure steps in this section are located on the drawer front panel for the channel being adjusted. IOP-002 d. PLACE the LEVEL TRIP switch in the BYPASS position.

e. VERIFY the LEVEL TRIP BYPASS indicator is ILLUMINATED.

f. VERIFY the SOURCE RANGE TRIP BYPASS status light on the RTGB is ILLUMINATED for the channel being adjusted.

g. VERIFY the NIS TRIP/DROP ROD BYPASS annunciator on the RTGB is ILLUMINATED (APP-005-D4) . h. RECORD the neutron level meter indication (Background).

N-31 CPS __ _ N-32CPS __ _ i. PLACE the HIGH FLUX AT SHUTDOWN switch in the BLOCK position.

j. VERIFY the HI FLUX AT SHUTDOWN ALARM BLOCK annunciator is ILLUMINATED on the RTGB (APP-005-B1).

Rev. 19 Page 15 of 221 Section 8.4.1 Page 3 of 5 8.4.1.2 (Continued) INIT INIT ! OP-002 k. CALCULATE the CPS neutron level setpoint [3 times the reading in Step 8.4.1.2.h OR per EST-048, EST-049, OR LP-551]. (ACR 95-00294) N-31 CPS __ N-32CPS __ I. PLACE the OPERATION SELECTOR Switch in the LEVEL ADJ position. N-31 N-32 m. CHECK APP-005-D3, NIS CHANNEL TEST annunciator on the RTGB, is ILLUMINATED. __ n. ADJUST LEVEL ADJ Pot until the CPS Neutron Level Meter indication is at the desired trip point in Step 8.4.1.2.k.

o. SLOWLY ADJUSTTRIPADJ on NC-103 until the Bistable just trips as indicated by the HIGH FLUX AT SHUTDOWN Lamp. (Adjustment inside drawer) p. IF the alignment can NOT be satisfactorily completed, THEN NOTIFY the SSO. q. TURN the LEVEL ADJUST to the full counter-clockwise (CCW) position.

Rev. 19 Page 16 of 221 ATTACHMENT 10.5 Page 2 of 3 RESETTING SR HI FLUX AT SHUTDOWN ALARM 3. Approval from the Manager -Operations, or designee, to insert both Shutdown Banks has been received.

4. IF Rods are withdrawn, THEN insert all Rods as follows [CAPR 173910-11]: 5. 6. 1) Select SBA on the Rod Bank Selector switch. 2) Insert the Shutdown Bank "A" to 005 steps. 3) Depress the REACTOR TRIP pushbutton.

Adjust the HI FLUX AT SHUTDOWN alarm to 3 times the AII-Rods-In count rate lAW OP-002. (ACR 95-00294) IF desired to withdraw the Shutdown Bank and Control Bank Rods, THEN perform the following [CAPR 173910-11]:

1) Perform the following while maintaining RCS temperature constant:

a. Verify closed the Reactor Trip Breakers.

b. Depress the ROD CONTROL STARTUP pushbutton.

c. Verify the Group Step Counters indicate 000. d. Check Individual Rod Position Indicators are within 7.5 inches of the Bank average rod height. e. Select SBA on the Rod Bank Selector switch. NOTE: Criticality shall be anticipated at any time, when the Shutdown Bank OR Control Bank Rods are being withdrawn.

f. Withdraw Shutdown Bank "A" to 225 steps. 2) Select SBB on the Rod Bank Selector switch. 3) Withdraw Shutdown Bank "B" to 005 steps. Rev. 103 Page 118 of 1241 HLC-08 NRC Written Exam 62. Given the following:

-The plant was operating at 100% RTP when a Small Break LOCA occurred. -Safety Injection has been actuated. -RCPs were tripped due to loss of subcooling. Which ONE (1) of the following describes how the S/Gs are controlled to ensure core cooling is maintained during this event? A. S/G levels are NOT required to be maintained at a minimum level due to SI injection flow. B. S/G levels are maintained at > 8% [18%] to provide additional heat removal capacity. C. S/G levels are NOT required as long as core exit thermocouples are maintained less than 700 OF. D. S/G AFW flow rates are maintained at 80-90 GPM to provide additional heat removal capacity. 62 HLC-08 NRC Written Exam 62. 035 A2.06 OOllSTEAM GENERATORJ2/2/4.5/4.6IROILOWININNEW -200S/PATH-I-005 Given the following: -The plant was operating at 100% RTP when a Small Sreak LOCA occurred. -Safety Injection has been actuated. -RCPs were tripped due to loss of subcooling. Which ONE (1) of the following describes how the S/Gs are controlled to ensure core cooling is maintained during this event? A. S/G levels are NOT required to be maintained at a minimum level due to SI injection flow. S/G levels are maintained at > 8% [18%] to provide additional heat removal capacity. C. S/G levels are NOT required as long as core exit thermocouples are maintained less than 700 of. D. S/G AFW flow rates are maintained at 80-90 GPM to provide additional heat removal capacity. The correct answer is S. A: Incorrect -Secondary Heat Sink is required as long as RCS pressure is greater than S/G pressure and RCS temperature is > 350 of. B: Correct -Maintaining S/G levels> 8% ensures Secondary Heat Sink is maintained as long as RCS pressure is greater than S/G pressure. C: Incorrect -Secondary Heat Sink is required as long as RCS pressure is greater than S/G pressure and RCS temperature is > 350 of. 0: Incorrect -AFW flow rates of 80-90 GPM maintains the S/G internals wet when ALL S/Gs are faulted and does NOT provide 300 GPM minimum flow. Exam Question Number: 62

Reference:

PATH-1 SO, Page 37; FRP-H.1 SO, Page 45. KA Statement: Ability to (a) predict the impacts of the following malfunctions or operations on the SG; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Small break LOCA. History: New -Written for HLC-08 NRC Exam. 69 GRID WOG BASIS/DIFFERENCES STEP E-9 SSD DETERMINATION This is an SSD per criterion

10. 28 RNP STEP ,., .. CONTROL AFW FLOW TO MAINTAIN S/G LEVELS BETWEEN 8% AND 50% 29 RNP STEP ANY S/G WITH UNCONTROLLED LEVEL INCREASE (with transition to Path-2) WOG BASIS ,. PURPOSE: To ensure adequate feed flow or SG inventory for secondary heat sink requirements BASIS: The minimum feed flow requirement satisfies the feed flow requirement of the Heat Sink Status Tree until level in at least one SG is restored into the narrow range. Narrow range level is reestablished in all SGs to maintain symmetric cooling of the RCS. The control range ensures adequate inventory with level readings on span. The transition to E-3, STEAM GENERATOR TUBE RUPTURE, responds to an increasing level which would be observed following a SGTR. KNOWLEDGE:

• "Level increase in an uncontrolled manner" means that the operator cannot control level using available equipment, i.e., level continues to rise even when all feed flow valves to that SG are fully closed.
• This step is a continuous action step. RNP DIFFERENCES/REASONS There are no significant differences.

SSD DETERMINATION This is not an SSD. RNP STEP R-19'S, R-31'S, AND R-15 RAD LEVELS NORMAL (with transition to Path-2) WOG BASIS PURPOSE: To identify any ruptured (failure in primary to secondary pressure boundary) SGs BASIS: Abnormal radiation in a steam generator indicates primary to secondary leakage. Since the air ejector and blowdown lines may have been isolated at the initiation of the transient, it may be necessary to check each steam generator at this time. Optimal recovery in dealing with a steam generator tube rupture is provided in E-3, STEAM GENERATOR TUBE RUPTURE. KNOWLEDGE:

• How to obtain secondary radiation level readings including signals that may need to be reset. * "Normal" means the value of a process parameter experienced during routine plant operations.

I PATH-1-BD Rev 18 Page 37 of 951 RNP waG BASISIDIFFERENCES STEP STEP 3 1 I FRP-H.1-BD RNP DIFFERENCES/REASONS The RNP procedure places the caution or note in an action step to prevent actions within cautions and noted as required by the writer's guide. SSD DETERMINATION This is an SSD per criterion

11. WOGBASIS BASIS: Before implementing actions to restore flow to the steam generators, the operator should check if secondary heat sink is required.

For larger LOCA break sizes, the RCS will depressurize below the intact steam generator pressures. The steam generators no longer function as a heat sink and the core decay heat is removed by the RCS break flow. For this range of LOCA break sizes, the secondary heat sink is not required and actions to restore secondary heat sink are not necessary. For these cases, the operator returns to the guideline and step in effect. Since Step 8 directs the operator to return to Step 1 if the loss of secondary heat sink parameters are not exceeded, break sizes that take longer to depressurize the RCS will be detected on subsequent passes through Step 1. If RCS temperature is low enough to place the RHR System in service, then the RHR System is an alternate heat sink to the secondary system. Therefore, an attempt is made, to place the RHR System in service in parallel to the attempts to reestablish feedwater flow. RCS pressure must be below normal RHR System pressure limits. KNOWLEDGE: The operator must be able to place the RHR System in service before the pressure and temperature limits are exceeded to make this alternate heat sink a valid option. Efforts to restore feedwater flow to the SGs should not be delayed if the RHR System is not a valid option. RNP DIFFERENCES/REASONS There are essentially no differences. SSD DETERMINATION This is not an SSD. Rev 22 Page 45 af 70 I HLC-08 NRC Written Exam 63. Given the following: -A Steam Generator Tube Rupture has occurred. -The crew is preparing to cooldown to a target temperature of 480 of lAW PATH-2. -A Loss of Off-Site Power occurs. -All equipment functions as required. Which ONE (1) of the following describes how the cooldown to target temperature will be accomplished? A. Intact S/G PORVs controllers set to 1005 PSIG. B. Steam Dumps in Pressure Control mode set to 3.95 (553 PSIG). C. Intact S/G PORVs controllers set to obtain full OPEN valve position. D. Steam Dumps in Pressure Control mode set to maximum rate. 63 HLC-08 NRC Written Exam 63.041 A4.06 OOllSTM DUMPffURB BYPASS/2/2/2.9/3.1IROIHIGWNINSALEM -2001IPATH-2-005 Given the following: -A Steam Generator Tube Rupture has occurred. -The crew is preparing to cooldown to a target temperature of 480 of lAW PATH-2. -A Loss of Off-Site Power occurs. -All equipment functions as required. Which ONE (1) of the following describes how the cooldown to target temperature will be accomplished? A. Intact SIG PORVs controllers set to 1005 PSIG. B. Steam Dumps in Pressure Control mode set to 3.95 (553 PSIG). C:I Intact SIG PORVs controllers set to obtain full OPEN valve position. D. Steam Dumps in Pressure Control mode set to maximum rate. The correct answer is C. A: Incorrect -Setting PORVs to 1005 PSIG will NOT cooldown the plant, but will maintain RCS temperature at 547 of. B: Incorrect -Condenser Steam Dumps are unavailable due to loss of Condenser from LOOP. Set at target temperature saturation pressure. C: Correct -Steam Line PORVs on intact S/Gs set to max rate. Condenser Steam Dumps are unavailable due to LOOP. 0: Incorrect -Steam Dumps unavailable due to loss of Condenser from LOOP. Exam Question Number: 63

Reference:

PATH-2 BD, Page 7. KA Statement: Ability to manually operate and/or monitor in the control room: Atmospheric relief valve controllers. History: Direct from Bank. 70

---

.-.. ------.-.---

... -.--.---... 2.2 __ ) principal systems/components affected by station blackout are the steam dump system, reactor coolant pumps (RCPs), and RCS pressure control. The effect of each of these on the system response and recovery is discussed. The steam dump system is designed to actuate following loss of load or reactor trip to limit the increase in secondary side pressure. Without offsite power available, the steam dump valves, which bypass the turbine to the condenser, will remain closed. Hence, energy transferred from the primary will rapidly increase steam generator pressures after reactor trip until the atmospheric relief valves lift to dissipate this energy, as shown in Figure 16. Since the secondary side temperature increase is greater, sensible energy transfer from the primary side following reactor trip is reduced. Consequently, RCS pressure decreases more slowly, as demonstrated in Figure 17, so that SI actuation and all attendant automatic actions, are delayed. A typical sequence of events without offsite power available is also presented in Table 1. RCPs trip on a loss of offsite power and a gradual transition to natural circulation flow ensues. The cold leg temperature trends toward the steam generator temperature as the fluid residence time in the tube region increases. Initially, the core delta-T decreases as core power decays following reactor trip and, subsequently, increases as natural circulation flow develops (Figures 18 and 19). Without RCPs running the upper head region becomes inactive and the fluid temperature in that region will Significantly lag the temperature in the active RCS regions. This creates a situation more prone to voiding during the subsequent cooldown and depressurization. Sufficient instrumentation and controls are provided to ensure that necessary recovery actions can be completed without offsite power available. Although the recovery method is the same with or without offsite power available, the equipment used may be different. Since automatic steam dump to the condenser is unavailable, reactor coolant temperatures may remain significantly greater than no-load temperature following reactor trip, as shown in Figure 20. Voiding may occur in the RCS as primary-to-secondary leakage depletes coolant inventory until manual actions to cool the RCS are initiated. In most cases, AFW flow will cool the RCS sufficiently to prevent voiding. This depends on the capacity of the AFW system and the actuation logic. For the results presented, AFW is assumed to actuate on the SI signal. J,!:le RCS is cooled using the power operated valyes (PORVs) on the intact steam generators since the dump valves nor the condenser would be without offsite power. Even with one steam generator out of service, these valves provide sufficient capacity to complete the initial RCS cooldown rapidly. Note that the hot leg temperature does not respond as quickly as the cold leg and steam generator temperatures since RCPs are not running. RCS pressure decreases rapidly during this cool down (Figure 21). The reactor coolant may saturate temporarily during the cooldown after the pressurizer empties (Figure 22). Under natural circulation conditions subsequent actions to isolate the affected steam generators and cool down the intact RCS loops may stagnate the affected loop. Consequently, the hot leg fluid in that loop may remain warmer than in the unaffected loops. Similarly, SI flow into the stagnant loop cold leg may rapidly decrease the fluid temperature in the cold leg, downcomer, and pump suction regions significantly below the rest of the RCS, as observed during the tube failure event at R.E. Ginna (Figure 23). With RCPs stopped, normal pressurizer spray would not be available. Consequently, RCS pressure must be controlled using pressurizer PORVs or auxiliary spray. Although a PORV enables more rapid RCS depressurization, it also results in an additional loss of reactor coolant which may rupture the pressurizer relief tank (PRT) and contaminate the containment. Auxiliary spray conserves reactor coolant but may create excessive thermal stresses in the spray nozzle which could result in nozzle failure. Consequently, it is recommended only if normal spray and pressurizer PORVs are not available. Since the upper head region is inactive, voiding may occur in this region during RCS depressurization. This will result in a rapidly increasing pressurizer level indication as water displaced from the upper head replaces steam released or condensed from the pressurizer. This behavior was observed during the Ginna tube failure event, as shown in Figure 24, when the pressurizer PORV failed to close. The extent of voiding is limited to the inactive regions of the RCS provided subcooling is maintained at the core exit. However, flashing in the inactive regions may slow further RCS depressurization to cold shutdown conditions. Even without offsite power, the E-3 recovery scheme establishes sufficient RCS subcooling, secondary side heat sink, and reactor coolant inventory to ensure SI flow is no longer required. The plant response is similar with or without offsite power available. Once SI flow is stopped, no additional primary-tosecondary leakage or uncontrolled radiological releases from the affected steam generators should occur. I PATH-2-BD Rev. 17 Page 7 of 110 I Question Number: Question: Answer: Justification: Tier/Group 1OCFR55.41 1OCFR55.43 Bank/New/

Modified KJA#: KJA Values: Cognitive Level:

References:

SALEM FOXTROT 2001 NRC WRITTEN EXAMINATION WORKSHEET Common 42 Unit 2 has sustained a Steam Generator Tube Rupture. The crew is preparing to cooldown to a target temperature of 497°F. A loss of Off-Site Power occurs. All equipment functions as required. Which one of the following describes how the cooldown to target temperature will be accomplished? A. Intact SG MS10s set to 25% B. Main Steam Dumps in Pressure Control mode set to 25 % C. Illtact SG MS 1 Os set to maximum rate D. Main Steam Dumps in Pressure Control mode set to maximum rate C MS10s on intact SGs set to max rate will not cause MSLI. Main steam dumps unavailable due to LOOP. C-9 unavailable with no CWPs. 25% put in there because that's the setpoint for normal steam dump cooldown for SGTR 112 41.10 New 038EA2.08 determine action to place unit in safe condition, no steam dumps available RO 3.8 SRO 4.4 Analysis EOP-SGTR-l, Step 15 LP SGTR-l, objective 3 HLC-08 NRC Written Exam 64. Which ONE (1) of the following describes how the Reactor Protection System provides Runback signals to the Turbine Generator? The RPS initiates a Turbine (1) runback when 2/3 OT Delta T (2) 2/3 OP Delta T signals reach their runback setpoints. A. (1) Reference (2) OR B. (1) Valve Limiter (2) OR C. (1) Reference (2) AND D. (1) Valve Limiter (2) AND 64 HLC-08 NRC Written Exam 64.045 K1.18 OOllMAIN TURBINE GEN/2/2/3.6/3.7fROfLOWfNINNEW -2008/RPS-006 Which ONE (1) of the following describes how the Reactor Protection System provides Runback signals to the Turbine Generator? The RPS initiates a Turbine (1) runback when 2/3 OT Delta T (2) 2/3 OP Delta T signals reach their runback setpoints. A'I (1) Reference (2) OR B. (1) Valve Limiter (2) OR C. (1) Reference (2) AND D. (1) Valve Limiter (2) AND The correct answer is A. A: Correct -Turbine Reference runbacks operate from 2/3 OT Delta T OR 2/3 OP Delta T on a 30 second timer. Runback signal is ON for 1.5 seconds and OFF for 28.5 seconds until runback signal clears. B: Incorrect -No valve limiter runback exists. C: Incorrect -Reference runback from 2/3 OT Delta T OR 2/3 OP Delta T, does NOT require BOTH. D: Incorrect -No valve limiter runback exists. Exam Question Number: 64

Reference:

SD-011, RPS, Page 23, Figure 15. KA Statement: Knowledge of the physical connections and/or cause-effect relationships between the MT/G system and the following systems: RPS. History: New -Written for HLC-08 NRC exam. 71 SD-011 REACTOR PROTECTION SYSTEM b. c. lout of 2 Intermediate Range Channels above 20 %, this block can be bypassed. 2 out of 3 OTAT's above calculated setpoint, this calculated setpoint is less than the OTAT Trip by 2%. Provided by the same bistable as the Turbine Runback. (See Figure 14) d. 2 out of 3 OPAT's above calculated setpoint, this calculated setpoint is less than the Trip by 2%. Provided by the same bistable as the Turbine Runback (See Figure 14) 4.4 Turbine Runbacks The Turbine Runback is designed to reduce Turbine Load and thus avoid an unnecessary trip. The Turbine Runback is caused by High Signals and are initiated by a Load Reference Reduction. 4.4.1 Load Reference Reduction (Figure 15) The Load Reference Reduction is caused when any of the following occur: RPS a. OT T -when 2 out of 3 Reactor Coolant Loop T' s (Th -Tc) exceeds the calculated setpoint. This causes a cyclic reduction in Turbine Load at the rate of 200%/minute. The cyclic duration is that of a 1.5 second runback (which equates to 5% Load Reduction) and a 28.5 second wait, if condition still exists, the 1.5 second runback and the 28.5 second wait repeats. This cyclic reduction continues until the OT T condition is corrected. The setpoint is identical to the Reactor Trip Setpoint except the Kl term becomes 1.1 065 2 % < the trip setpoint (Refer to 4.1.5.5).

b.

-when 2 out of 3 Reactor Coolant Loop (Th -Tc) exceeds the calculated setpoint. This causes a cyclic reduction in Turbine Load at the rate of 200%/minute. The cyclic duration is that of a 1.5 second runback (which equates to 5% Load Reduction) and a 28.5 second wait, if the condition still exists, the 1.5 second runback and the 28.5 second wait repeats. This cyclic reduction continues until the condition is corrected. Page 23 of30 Revision 8 INFORMATION USE ONLY TURBINE LOAD REFERENCE REDUCTIONS RPS-FIGURE-15 HIGH .HIGH OVER-OVER; TEMPERATURE POWER I D A . RELAY LOGIC .(CYCLIC) ON f4-28.5 SEC ...... t+-I.5SEC REDUCE LOAD REFERENCE AT 200% MIN RPSF15 HLC-08 NRC Written Exam 65. Given the following: -EPP-1, LOSS OF ALL AC POWER, is being performed. -The RCS has been isolated. -The Inside AO and maintenance technicians are working on starting an EOG. -There is NO SI signal present or required. -EOG "A" is finally started. Which ONE (1) of the following actions is necessary following the energization of 480V Bus E-1? Verify Service Water Pumps ... A. "A" and "B" are running. B. "A" and "C" are running. C. "B" and "0" are running. D. "C" and "0" are running. 65 \'--...-., HLC-08 NRC Written Exam 65.075 K2.03 OOl/CIRCULATING WATER/2/2/2.6/2.7IROIHIGH/NIAIRNP AUDIT -200l/SW-009 Given the following: -EPP-1, LOSS OF ALL AC POWER, is being performed. -The RCS has been isolated. -The Inside AO and maintenance technicians are working on starting an EDG. -There is NO SI signal present or required. -EOG "A" is finally started. Which ONE (1) of the following actions is necessary following the energization of 480V Bus E-1 ? Verify Service Water Pumps ... A'I "A" and "B" are running. B. "A" and "C" are running. C. "8" and "0" are running. D. "C" and "0" are running. The correct answer is A. A: Correct -"A" and "B" now have power available and should have started. B: Incorrect -"A" now has power available and should have started. "c" does NOT have power. C: Incorrect -"8" now has power available and should have started. "0" does NOT have power. 0: Incorrect -"c" and "0" do NOT have power. Exam Question Number: 65

Reference:

EOP-002, Page 11; SO-004, SW, Page 21. KA Statement: Knowledge of bus power supplies to the following: Emergency/essential SWS pumps. History: Direct from Bank. 72 8.0 * ** 480V-E1{ TC "480V-E1" \f C \l "1" } 480V*E1 Section 8.0 Page 1 of 1 POWER SUPPLY: NORMAL -4160V BUS 2 (52/13)LOCATION: E-1/E-2 ROOM CMPT LOAD TITLE CWD BKR NO. EDBS LOAD TAG NO. NO. EDBSNO. 17A PT'S & METERING EQUIPMENT (*) N/A N/A N/A 17B EMERGENCY DIESEL GENERATOR A TO 480V BUS E*1 890 52117B 480V-E1 18A PT'S & METERING EQUIPMENT (**) N/A N/A N/A 18B STATION SERVICE TRANSFORMER 2F TO 480V BUS E*1 892 52/18B 480V-E1 19A CV SPRAY PUMP A 287 52119A CV-SPRAY-PMP-A 19B CV RECIRC FAN, HVH-1 511 52119B HVH-1 19C SERVICE WATER PUMP B 832 52119C SW-PMP-B 20A AUX FEEDWATER PUMP A 651 52120A AFW-PMP-A 20B SERVICE WATER PUMP A 831 52/20B SW-PMP-A 20C CV RECIRC FAN, HVH-2 512 52120C HVH-2 21A FEED TO MCC*5 (NORM POWER) & MCC-16 1187 52121 A MCC-5, MCC-16 21B CHARGING PUMP B 162B 52/21 B CHG-PMP-B 21C SAFETY INJECTION PUMP A 237 52/21C SI-PMP-A 22A RESIDUAL HEAT REMOVAL PUMP A 214 52/22A RHR-PMP-A 22B 480V BUS E*1 SUPPLY TO SI PUMP B 891 52122B 480V-E1, E2 22C COMPONENT COOLING WATER PUMP B 205 52122C CCW-PMP-B Compartment 17 A also contains two amp meters, two amp meter switches, one volt meter, one volt meter switch, two undervoltage relays, four overcurrent relays, and two auxiliary relays. Compartment 18A also contains eight run time meters, three degraded grid relays, one degraded grid trip signal, three test switches, and one degraded grid voltage switch. I EDP-002 Rev. 10 Page 11 of 13\ SD-004 SERVICE WATER SYSTEM 5.5 5.6 sw Pump D power from its normal supply (E-2) to its emergency supply (DS Bus) by use of a Kirk-key operated breaker (located in the CCW Pump Room). Power Supplies Component SWPumpA -1 , SW Pump B -1 c. SWPump C E-d. SWPumpD E-2/ DS Bus e. SW Booster Pump A MCC-16 f. SW Booster Pump B MCC-18 g. V6-12A (South Header Isolation) MCC-5 h. V6-12B (Header Cross-Connect) MCC-5 i. V6-12C (Header Cross-Connect) MCC-6 j. V6-12D (North Header Isolation) MCC-6/5 k. V6-16A (North Header Turbine Bldg Isolation) MCC-9 1. V6-16B (South Header Turbine Bldg Isolation) MCC-lO m. V6-16C (Turbine Bldg header Isolation) MCC-1O/9 Electrical Manhole Sump Pumps The Service Water Pump Power cables are routed between the Intake Structure and the Reactor Auxiliary Building (RAB) through two (2) underground trenches. The cables enter the trenches though 2 Manholes between the RAB and the Radioactive Waste Building and the cables exit the trenches though 2 Manholes at the intake structure. Due to the high water table at the H.B. Robinson Plant the manholes are constantly being filled with water. Permanent sump pumps in Manholes M35 and M36 have been installed in order to improve the life span and reliability of the cables installed in these manholes (ESR 98-00319). The sump pumps have level switches installed in the manholes to provide automatic start of the pumps when water accumulates in the manhole sump. A sump pump control panel has been installed in the area of the manholes and has a manual pump switch installed to provide for manual operation of the pump. The control panel also has a pump run light for each pump that will be illuminated when the pumps are in operation. A level switch has been installed in the manholes to energize an alarm light and horn on the control panel if the water level in either of the manholes rises above 18 inches. The discharge piping of the sump pumps has been installed with freeze protection cable and insulation to ensure that the piping does not freeze during cold weather. The control panel includes a freeze protection test switch for each of the discharge lines freeze protection circuits and an ammeter to provide indication of the circuits' current draw. This modification also installed three (3) spare cables from PP-61 to the new sump pump control panel. A storage box has Page 21 of35 Revision 11 INFORMATION USE ONLY 1 . 055 G2.4.30 001/////1/1 QUESTIONS REPORT for AUDIT Given the following conditions:

• EPP-001, Loss of All AC Power, is being performed.
• The RCS has been isolated.
• The Inside AO and maintenance are working on starting an EOG.
• There is no SI signal present or required.

The "A" EOG is finally started. Which ONE (1) of the following is the first action to be taken following the energization of Bus E-1? A'I Start "A" and "BII SW Pumps as necessary to obtain SW pressure of at least 40 psig B. Start IIAII and IIBII SW Pumps as necessary to obtain SW pressure of at least 50 psig C. Start IICII and 110 11 SW Pumps as necessary to obtain SW pressure of at least 40 psig O. Start IICII and 110" SW Pumps as necessary to obtain SW pressure of at least 50 psig A is correct. Need SW to support EOG. IIA" and IIBII now have power available. Common Question 050 Tier 1 Group 1 KIA Importance Rating -RO 3.2/ SRO 3.2 Emergency Procedures / Plan Knowledge of which events related to system operations/status should be reported to outside agencies. Reference(s) -EPP-1 Proposed References to be provided to applicants during examination -None Learning Objective -Question Source -Bank Question History -2004 Harris #12 Question Cognitive Level -Comprehension 10 CFR Part 55 Content -41 Comments -Needs KA Change Category 1: Category 3: Category 5: Category 7: Tuesday, June 17, 2008 6:49:52 PM Category 2: Category 4: Category 6: Category 8: 1 HLC-08 NRC Written Exam 66. Given the following: -Operators are performing valve lineups to support BOP Flush. -Several valves above Heater Drain Tank "B" are to be repositioned. -The valves are approximately 15 feet above the floor level. What fall protection or safety measures is required to perform the valve manipulations safely? A. Tie or secure a ladder to structural components at BOTH the top and bottom, use 3-point contact while on the ladder. B. Don a full body harness and attach it to the Extraction Steam line support or snubber rods. C. Don a Bosun's Belt with a 6 foot lanyard attached to an approved anchorage point. D. Don a full body harness attached to an approved anchorage point. 66 HLC-08 NRC Written Exam 66. G2.1.26 OOllEQUIPMENT CONTROLl3/3.4/ROILOW/N/A/NEW -200S/GET-FALLPROT Given the following: -Operators are performing valve lineups to support BOP Flush. -Several valves above Heater Drain Tank "B" are to be repositioned. -The valves are approximately 15 feet above the floor level. What fall protection or safety measures is required to perform the valve manipulations safely? A. Tie or secure a ladder to structural components at BOTH the top and bottom, use 3-point contact while on the ladder. B. Don a full body harness and attach it to the Extraction Steam line support or snubber rods. C. Don a Bosun's Belt with a 6 foot lanyard attached to an approved anchorage pOint. Don a full body harness attached to an approved anchorage point. The correct answer is D. A: Incorrect -Procedure requires that extension ladder be attached to structural components at the top of the ladder while the ladder is secured at the bottom by an individual. B: Incorrect -Fall protection is to be attached to engineered anchorage pOints. Line or snubber supports are NOT approved anchorage pOints. C: Incorrect -Bosun's Belts are NOT approved fall protection at Robinson. D: Correct -Full body harness properly used and attached to approved anchorage points is required to safely perform work. Exam Question Number: 66

Reference:

SAF-NGGC-2172, Sect. 9.17. KA Statement: Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen). History: New -Written for HLC-08 NRC Exam. 73 9.16 Safe Sling Operations 9.16.1 Slings shall be inspected, and used in accordance with SAF-SUBS-00028 or site rigging procedures. 9.16.2 Damaged or defective slings shall not be used. 9.16.3 Slings shall not be shortened with knots, bolts or other makeshift devices. 9.16.4 Slings shall not be kinked. 9.16.5 Loads lifted in a basket hitch shall be balanced to prevent spillage. 9.16.6 Slings shall be securely attached to the load. 9.16.7 Suspended loads shall be kept clear of all obstructions. 9.16.8 Personnel shall be kept clear of suspended loads and loads about to be lifted. 9.16.9 Hands and fingers shall not be placed between sling and load while sling is being tightened around load. 9.16.10 Tag lines shall be employed to guide loads and prevent excessive sway during load movement. 9.17 Fall Protection 9.17.1 Fall protection should be used in accordance with SAF-ESGX-00002, SAF-SUBS-00019, SAF-SUBS-00021 and site procedures. 9.17.2 All fall protection users must be properly trained and qualified as directed in SAF-ESGX-00002. 9.17.2 Safety harnesses and lanyards shall be inspected by each user prior to use. 9.17.3 Full body harnesses with shock absorbing lanyards shall be used when working at unprotected elevations above six feet. 9.17.4 Lanyards shall not be tied to themselves in a choked manner. If the anchorage point is too large for the lanyard to hook directly on, the employee shall use an anchorage strap or other approved attachment device to attach to the anchorage point. Note: Lanyards with an installed O-Ring can be connected back to the O-Ring. 9.17.5 Escorts of contractors and vendors shall ensure fall protection requirements are followed. I SAF-NGGC-2172 REV. 9 Page 19 of 53\ 9.17 Fall Protection 9.17.6 Occupied elevated fixed ladder access points and floor openings shall be guarded by a floor cover, guardrail or other substantial physical barrier, or shall be constantly attended to prevent personnel from falling into openings. --7 9.17.7 Employees shall wear properly tied-off safety harnesses whenever working in man lifts or aerial lifted baskets. It is not required for employees to tie-off in manlifts that move only in the vertical direction and have OSHA approved handrails, midrails and toeboards. 9.17.8 A safety harness is not to be used to support any employee. It is a back-up safety device only in case the employee slips or falls. 9.17.9 When utilizing personal fall protection, employees shall select appropriate anchorage points as indicated in Attachment 2 or Attachment 3 in SAF-ESGX-00002. 9.17.10 Climbing on insulated pipes, pipe hanger components, and instrument lines is prohibited, except where the nature of the work is such that no damage is expected and other options are determined as not being feasible. 9.17.11 Climbing on cable trays is to be minimized. In cases where access to equipment can ONLY be attained by climbing on or through cable trays, the access will be permitted while exercising maximum caution. The possible damage to insulation, instrument lines, and power cables make strict compliance with this rule mandatory. 9.18 Ladders 9.18.1 Ladders should be erected, used and stored in accordance with SAF-SUBS-00019, SAF-SUBS-00022 and SAF-SUBS-00036 or site procedures. 9.18.2 Ladders shall be inspected in accordance with the site's maintenance inspection programs AND prior to each use by the ladder user. At a minimum, the following items should be included in the inspection:

• Missing or damaged components (e.g., cracked side rails, bent or damaged rungs, missing non-slip material on feet of ladder, loose or missing rivets, screws, or bolts, etc.)
• All fasteners are secure
• All moving parts are in good working order
• Side rails and rungs are free from dirt, oil or other foreign material.

Defective ladders shall be removed from service for repair or destruction and tagged or marked, "Dangerous, Do Not Use" or other suitable verbiage. 9.18.3 Step ladders shall be used appropriately with legs fully extended and locked, and shall not be used as extension ladders. SAF-NGGC-2172 REV. 9 Page 20 of 53 I ES-401, Rev. 9 Written Examination Review Worksheet -RO Form ES-401-9 1. 2. 3. Psychometric Flaws 4. Job Content Flaws 5. Other 6. 7. Q# LOK LOD (F/H) (1-5) Stem Cues T/F Credo Partial Job-Minutia #/ Back-Q= SRO u/E/S Explanation Focus Dis!. Link units ward KIA Only Instructions [Refer to Section 0 of ES-401 and Appendix B for additional information regarding each of the following concepts.]

1. Enter the level of knowledge (LOK) of each question as either (F)undamental or (H)igher cognitive level. 2. Enter the level of difficulty (LOD) of each question using a 1 -5 (easy -difficult) rating scale (questions in the 2 -4 range are acceptable).

3. Check the appropriate box if a psychometric flaw is identified:

• The stem lacks sufficient focus to elicit the correct answer (e.g., unclear intent, more information is needed, or too much needless information).
• The stem or distractors contain cues (Le., clues, specific determiners, phrasing, length, etc).
• The answer choices are a collection of unrelated true/false statements.
• The distractors are not credible; single implausible distractors should be repaired, more than one is unacceptable.
• One or more distractors is (are) partially correct (e.g., if the applicant can make unstated assumptions that are not contradicted by stem). 4. Check the appropriate box if a job content error is identified:
• The question is not linked to the job requirements (Le., the question has a valid KIA but, as written, is not operational in content).
• The question requires the recall of knowledge that is too specific for the closed reference test mode (Le., it is not required to be known from memory).
• The question contains data with an unrealistic level of accuracy or inconsistent units (e.g., panel meter in percent with question in gallons).
• The question requires reverse logic or application compared to the job requirements.

5. Check questions that are sampled for conformance with the approved KIA and those that are designated SRO-only (KIA and license level mismatches are unacceptable

). 6. Based on the reviewer'S judgment, is the question as written (U)nsatisfactory (requiring repair or replacement), in need of (E)ditorial enhancement, or (S)atisfactory?

7. At a minimum, explain any "U" ratings (e.g., how the Appendix B psychometric attributes are not being met). 1 X S? Should not use the word approximately.

Show how you determine the points used. 2 U A & B are not plausible. Are there any scenario, where an automatic valve alignment would occur while an operator is performing a manual transfer or alignment on that system? 3 1 U LOD. This question is very simple. What indication are provided tq indicate there might be a failure of the #2 seal? 4 2 S ... ----

1. 2. 3. Psychometric Flaws 4. Job Content Flaws 5. Other 6. 7. Q# LOK LOD (F/H) (1-5) Stem Cues T/F Credo Partial Job-Minutia #1 Back-Q= SRO UlE/S Explanation Focus Dis!. Link units ward KIA Only 5 1 U Easy question.

Fluctuating -cycling. All of the other distractors would result in constant flow or no flow. Very little system knowledge is needed to answer this question. 6 X S/E Crew "is"-...... Path-1 and is unable ....... 7 2 S 8 2 X E In distractor C you are increasing the possibility of a release. Why would this be plausible? 9 2 S 10 X X U Look at question 7. Very little difference in the two questions 11 2 S 12 X X U The stem says that "Both PAM operable." There is no need for the applicant to say that he must verify PAM is operable as stated in the distractors A & B. Two implausible distractors A & B. 13 H 2 S 14 2 H S 15 F 2 S 16 1 S This is a memory level question. Could be improved. 17 F 2 S 18 F 2 S 19 H 1/2 U Based on the information provided, limited information is needed to answer the question -IRPI indications increasing and Tavg 1.5 degrees higher than Tref and rods in AUTO = UNC ROD withdrawal. Distractors are not plausible. As written LOD . . ... . .....

1. 2. 3. Psychometric Flaws 4. Job Content Flaws 5. Other 6. 7. Q# LOK LOD (F/H) (1-5) Stem Cues T/F Credo Partial Job-Minutia #/ Back-Q= SRO UtE/S Explanation Focus Dist. Link units ward KIA Only 20 X U There may be two correct answer or the answer identified as correct is not. If there is a rod that failed to insert, some actions must be taken to insert that rod as some point. The answer you provided as correct states "No actions required." 21 H 2 S 22 2 X E/S Need to look at plausibility of distractor A 23 F? 2 X E Distractor A is not plausible 24 F 2 S 25 F? 1 U LOD. The question meets the KIA. Please explain the operational value of this question.

How often and in what procedures require the RO to make this determination/perform such a calculation? 26 S 27 X U Distractors A & B are not plausible 28 S 29 X X U? Distractor B is not plausible. You said that it was isolated. KIA not matched 30 S 31 3 S 32 2 S 33 2 S 34 2 S 35 2 S 36 1 X U Distractor A is not plausible. Distractor B is not plausible. Distractors , not related to question asked. ----

1. 2. 3. Psychometric Flaws 4. Job Content Flaws 5. Other 6. 7. Q# LOK LOD (F/H) (1-5) Stem Cues T/F Credo Partial Job-Minutia #1 Back-Q= SRO utE/S Explanation Focus Dis!. Link units ward KIA Only 37 112 X E LOD. In distractor C& 0 are you attempting to say auto start? If so, why not use Auto-start/automatically started? Distractor 0 plausible?

Would such a condition exist for the given plant equipment lineup? Maintenance is in progress -if proper safety precautions are in place, precautions would be in place to prevent equipment from starting -personnel safety is one of the first things addressed when working on equipment -that means making sure that equipment will not auto start ...... 38 2 X S/E Distractors are weak. All systems functioned as designed -please explain why anyone would expect the a design limit to be exceed. 39 2 S 40 2 S 41 1 LOD To answer this question all one need to remember is a required flow of 300 GPM is needed. The 300 GPM flow requirement is considered common knowledge. Flow is increased by opening a valve/valves. 42 3 S 43 3 S 44 2 S 45 1 X LOD. Easy distractors. I am not sure if you can do much with this question. Is there a valve that can not be closed form the Panel???? I 46 X U As written the question has two correct answers (A & B) ! 47 2 S Easy 48 X U Two correct answers (A & B would solve the problem). 49 2 X U Two correct answers -A & 0 50 1/2 S Easy question

1. 2. 3. Psychometric Flaws 4. Job Content Flaws 5. Other 6. 7. Q# LOK LOD (F/H) (1-5) Stem Cues T/F Credo Partial Job-Minutia #1 Back-Q= SRO utE/S Explanation Focus Dis!. Link units ward KIA Only 51 2 X S? Could C also be correct? 52 2 S 53 1 X ? U LOD. We know that SG A is isolated. A & C removed immediately.

Distractors do not appear to be plausible. Make sure KIA matches. 54 X U? Please explain why you consider this a KIA match 55 1 X U LOD. KIA not matched. Setpoint question. As written there is more than one correct answer. 56 2 S 57 2 S 58 2 S Question could be improved!! 59 1 X X U What are you asking? The stem says that Train A Plasma display is inop ..... Then use Train B. Then you ask how to determine subcooling and CET -answer CETC temperatures from Train A and B CETCs. Distractors A & D are not plausible. Are temperatures gotten from a *unction box? Why would one be required to look at reading from only one train to make a key determination? 60 2 S ;'vt ttJ)E P!f:/.t'V}Lf-PS: /;!/ 57e>y} 61 2 S L 114J'!l(;.e. 'C>--5/eM "'-'uA-62 X U Distractors A & D do not appear to be plausible ._ C \,4 .t\ 10 ../ 63 2 S /Ii If}; If Il C If If J.J IS is h tfilJ prZ71; (11) 64 2 S Dk L-b/ 65 2 S M f\.\) E:.. C *\"1 ;q Iv & r-J:i;;. [ oft::.) ST.E.M ., 66 1 X E Distractor C is not plausible. What are Bosun's used for? 67 2 S? See if this question is on the SRO exam. -;I>;rr74.J -;-t'P//,? , fllvJ> i.iXJ --0 :k.

1. 2. 3. Psychometric Flaws 4. Job Content Flaws 5. Other 6. 7. Q# LOK LOD (F/H) (1-5) Stem Cues T/F Credo Partial Job-Minutia #/ Back-Q= SRO U/E/S Explanation Focus Dist. Link units ward KIA Only 68 1 X U/E? LOD Do you think this question addresses the knowledge of a /, L. process? tyt5!t, /9 7 ,ot!P1:l!:E!)U;4f->/ 69 U LOD It appears that this question can be answered by answering which system/component is most important to plant safety/operation.

70 1 U LOD General rad worker question. Is this specifically RO knowledge? 71 1 ? LOD 72 1 ? LOD 73 1 X ? LOD Simple memory -Distractor C. 74 1 ? LOD 75 X X U "A" could be correct. As written the stem states that all immediate actions have been completed. Therefore, is it not right to say that PATH-1 actions have been completed? Have you not already ensured that the reactor and turbine ......... ? If only one train of SI and RHR is required, then there may be not correct answer. In your reference you did not provide the bases. Distractors C & D are not plausible. When is an operator required to wait a period of time before attempting to start/make happen an automatic action once it was observed as not having occurred? I I ES-401, Rev. 9 Written Examination Review Worksheet -SRO Form ES-401-9 7. Q#ILOK LOD

1. I 2. (F/H) (1-5) Explanation Focus Dis!. Link units I ward I KIA IOnly Instructions

[Refer to Section D of ES-401 and Appendix B for additional information regarding each of the following concepts.]

1. Enter the level of knowledge (LOK) of each question as either (F)undamental or (H)igher cognitive level. 2. Enter the level of difficulty (LOD) of each question using a 1 -5 (easy -difficult) rating scale (questions in the 2 -4 range are acceptable).

3. Check the appropriate box if a psychometric flaw is identified:

The stem lacks sufficient focus to elicit the correct answer (e.g., unclear intent, more information is needed, or too much needless information). The stem or distractors contain cues (i.e., clues, specific determiners, phrasing, length, etc). The answer choices are a collection of unrelated true/false statements. The distractors are not credible; single implausible distractors should be repaired, more than one is unacceptable. One or more distractors is (are) partially correct (e.g., if the applicant can make unstated assumptions that are not contradicted by stem). 4. Check the appropriate box if a job content error is identified: The question is not linked to the job requirements (i.e., the question has a valid KIA but, as written, is not operational in content). The question requires the recall of knowledge that is too specific for the closed reference test mode (i.e., it is not required to be known from memory). The question contains data with an unrealistic level of accuracy or inconsistent units (e.g., panel meter in percent with question in gallons). The question requires reverse logic or application compared to the job requirements.

5. Check questions that are sampled for conformance with the approved KIA and those that are designated SRO-only (KIA and license level mismatches are unacceptable

). 6. Based on the reviewer's judgment, is the question as written (U)nsatisfactory (requiring repair or replacement), in need of (E)ditorial enhancement, or (S)atisfactory?

7. At a minimum, explain any HUH ratings (e.g., how the Appendix B psychometric attributes are not being met). 76 x x x U There appears to be unnecessary information in the stem. You identify procedures, then ask what procedures should be used. I do not see this as a SRO only question.

If the controller was in AUTO, and the RO noticed that the controller was operating erratically, he/she would take manual control, realize what procedure should be entered, perform immediate actions from memory to correct the problem based on plant conditions -SRO directionsfinstructions would not be required to assure that the immediate actions are complete.

1. 2. 3. Psychometric Flaws 4. Job Content Flaws 5. Other 6. 7. Q# LOK LOD (F/H) (1-5) Stem Cues T/F Credo Partial Job-Minutia #1 Back-Q= SRO U/E/S Explanation Focus Dist. Link units ward KIA Only 77 X U Distractor 0 is not plausible.

There is no information in the stem which H would indicate that a loss of IA had occurred. If we allow the applicant to make that assumption, then the answer could be correct if the failed close position is closed. Assumptions should not be made in selecting the answer. How can Oistractor A be plausible, if you say"no action required" then conclude with saying that you must "Ensure ...... " This would be an action 78 H X X U This question can be answered with system knowledge only. Oistractor 0 is not plausible. If a pump is cavitating, why would anyone think that reducing flow would solve the problem. 79 H S 80 H X X X X U Two correct answers -C & O. Based on the information given and the current plant conditions, why would anyone think that they would exit LOSS of instrument air? The stem asked what procedure is required to restore cool cooling -you only need to remember what procedure number. Do reactor operators not know this? Explain why you consider this SRO only. 81 F S/E/U Who is required to know the bases of the cautions in procedures. Can ? this question be answered based on system knowledge (how system operate for specific conditions)? The basis states "The Caution is provided to warn the Operator of the possibility of equipment performing uncontrolled starts." 82 H 2 X X U Not SRO only. System question that an RO can answer. Distractor D not plausible. 83 H X S/E Take a look at distractor B. Make sure that it is not correct. Why do you consider distractor 0 plausible? 84 F 2 X U Distractors C & 0 are not plausible. Why do you consider A plausible. The reactor is shut down -Why would one consider Reactor Core Safety Limits a concern when in FRCP-C.1?

---

1. 2. 3. Psychometric Flaws 4. Job Content Flaws 5. Other 6. 7. Q# LOK LOO (F/H) (1-5) Stem Cues T/F Credo Partial Job-Minutia #/ Back-Q= SRO U/E/S Explanation Focus Oist. Link units ward KIA Only 85 H X ? U What is this question asking? It appears that the question is asking which safety limit is affected once a cool-down is started while implementing FRP-C.1. (Pressure

/ Temperature) This is RO knowledge, in that they know that pressure and temperature limits are of a concern. They may not know the TS number, but they can identify the limit. Oistractors A & 0 are not plausible. 86 H X U Not an SRO question. System knowledge question. For the conditions given, after the standby pump is started, observations noted by the RO after starting the pump. He/she recognizes that pressure is increasing and takes appropriate actions according to APP. Why would one think there is a malfunction of the RCS pressure control when we said that the speed controller for the pump was set at maximum. Is the SRO really providing directions or is the RO performing actions and the SRO is agreeing? Would the RO not respond by stopping the activities that started the pressure increase ..... Stopping the standby pump .... ? 87 H This is an RO question. To answer the question, based on information X U in the stem and the way the distractors are written, all you need to realize is the fact that an SI did not occur. and the distractors, all you need to know to answer this question is ...... Is SI required, yes -go to Path -1. 88 F X U Please explain why you consider this a SRO only question. 89 H X U As written this is a systems question requiring only RO knowledge to answer. 90 F 1 U LOO 91 H S? Need to make sure there are not two correct answers (0 & B) 92 H S? May be RO Knowledge. Based on the information provided could an RO not answer the question once it is concluded that RCS leakage is occurring. You could leave on the procedure number and selected correct answer based on name of procedure only Excessive Leakage. 93 F 1 U LOO. Identify a system where oxygen limits is required to be maintained to prevent corrossion . , .... ---....... ----

1. 2. 3. Psychometric Flaws 4. Job Content Flaws 5. Other 6. 7. Q# LOK LOD (F/H) (1-5) Stem Cues T/F Credo Partial Job-Minutia #1 Back-Q= SRO U/E/S Explanation Focus Dist. Link units ward KIA Only 94 F S? Could C be correct? 95 F? X U? Not SRO only. Who makes the entry? 96 F 1 U LOD. No knowledge of procedures is required.

Distractors B not plausible. Distractor A could be correct. 97 F 1 X ? U Distractor C & D are not plausible. 98 H 3 S 99 F 1 LOD Common knowledge, RO would know this, but it is the SRO responsibility to know what to do given the conditions. Memory question .... Which AOPs are considered concurrent AOPs? Which AOPs should be performed concurrently while performing procedures in the EOP network? 100 F 1 X X E Could distractor C be correct? Is knowledge of strategy of actions mean to describe the bases? I I ES-401, Rev. 9 Written Examination Review Worksheet -SRO Form ES-401-9 1. 2. 3. Psychometric Flaws 4. Job Content Flaws 5. Other 6. 7. Explanation Focus Dist. Link units I ward I KIA IOnly Instructions [Refer to Section 0 of ES-401 and Appendix B for additional information regarding each of the following concepts.]

1. Enter the level of knowledge (LOK) of each question as either (F)undamental or (H)igher cognitive level. 2. Enter the level of difficulty (LOO) of each question using a 1 -5 (easy -difficult) rating scale (questions in the 2 -4 range are acceptable).

3. Check the appropriate box if a psychometric flaw is identified:

The stem lacks sufficient focus to elicit the correct answer (e.g., unclear intent, more information is needed, or too much needless information). The stem or distractors contain cues (Le., clues, specific determiners, phrasing, length, etc). The answer choices are a collection of unrelated true/false statements. The distractors are not credible; single implausible distractors should be repaired, more than one is unacceptable. One or more distractors is (are) partially correct (e.g., if the applicant can make unstated assumptions that are not contradicted by stem). 4. Check the appropriate box if a job content error is identified: The question is not linked to the job requirements (Le., the question has a valid KIA but, as written, is not operational in content). The question requires the recall of knowledge that is too specific for the closed reference test mode (Le., it is not required to be known from memory). The question contains data with an unrealistic level of accuracy or inconsistent units (e.g., panel meter in percent with question in gallons). The question requires reverse logic or application compared to the job requirements.

5. Check questions that are sampled for conformance with the approved KIA and those that are designated SRO-only (KIA and license level mismatches are unacceptable).

6. Based on the reviewer's judgment, is the question as written (U)nsatisfactory (requiring repair or replacement), in need of (E)ditorial enhancement, or (S)atisfactory?

7. At a minimum, explain any "U" ratings (e.g., how the Appendix B psychometric attributes are not being met). 76 x x x U There appears to be unnecessary information in the stem. You identify procedures, then ask what procedures should be used. I do not see this as a SRO only question.

If the controller was in AUTO, and the RO noticed that the controller was operating erratically, he/she would take manual control, realize what procedure should be entered, perform immediate actions from memory to correct the problem based on plant conditions -SRO directionslinstructions would not be required to assure that the immediate actions are complete.

1. 2. 3. Psychometric Flaws 4. Job Content Flaws 5. Other 6. 7. Q# LOK LOO (F/H) (1-5) Stem Cues T/F Credo Partial Job-Minutia #1 Back-Q= SRO utE/S Explanation Focus Oist. Link units ward KIA Only 77 X U Oistractor 0 is not plausible.

There is no information in the stem which H would indicate that a loss of IA had occurred. If we allow the applicant to make that assumption, then the answer could be correct if the failed close position is closed. Assumptions should not be made in selecting the answer. How can Oistractor A be plausible, if you say"no action required" then conclude with saying that you must "Ensure"",," This would be an action 78 H X X U This question can be answered with system knowledge only. Distractor 0 is not plausible. If a pump is cavitating, why would anyone think that reducing flow would solve the problem. 79 H S 80 H X X X X U Two correct answers -C & O. Based on the information given and the current plant conditions, why would anyone think that they would exit LOSS of instrument air? The stem asked what procedure is required to restore cool cooling -you only need to remember what procedure number. Do reactor operators not know this? Explain why you consider this SRO only. 81 F S/E/U Who is required to know the bases of the cautions in procedures. Can ? this question be answered based on system knowledge (how system operate for specific conditions)? The basis states "The Caution is provided to warn the Operator of the possibility of equipment performing uncontrolled starts." 82 H 2 X X U Not SRO only. System question that an RO can answer. Oistractor 0 not plausible. 83 H X S/E Take a look at distractor B. Make sure that it is not correct. Why do you consider distractor 0 plausible? 84 F 2 X U Oistractors C & 0 are not plausible. Why do you consider A plausible. The reactor is shut down -Why would one consider Reactor Core Safety Limits a concern when in FRCP-C.1?

1. 2. 3. Psychometric Flaws 4. Job Content Flaws 5. Other 6. 7. Q# LOK LOO (F/H) (1-5) Stem Cues T/F Credo Partial Job-Minutia #/ Back-Q= SRO u/E/S Explanation Focus Oist. Link units ward KIA Only 85 H X ? U What is this question asking? It appears that the question is asking which safety limit is affected once a cool-down is started while implementing FRP-C.1. (Pressure

/ Temperature) This is RO knowledge, in that they know that pressure and temperature limits are of a concern. They may not know the TS number, but they can identify the limit. Oistractors A & 0 are not plausible. 86 H X U Not an SRO question. System knowledge question. For the conditions given, after the standby pump is started, observations noted by the RO after starting the pump. He/she recognizes that pressure is increasing and takes appropriate actions according to APP. Why would one think there is a malfunction of the RCS pressure control when we said that the speed controller for the pump was set at maximum. Is the SRO really providing directions or is the RO performing actions and the SRO is agreeing? Would the RO not respond by stopping the activities that started the pressure increase ..... Stopping the standby pump .... ? 87 H This is an RO question. To answer the question, based on information X U in the stem and the way the distractors are written, all you need to realize is the fact that an SI did not occur. and the distractors, all you need to know to answer this question is ...... Is SI required, yes -go to Path -1. 88 F X U Please explain why you consider this a SRO only question. 89 H X U As written this is a systems question requiring only RO knowledge to answer. 90 F 1 U LOO 91 H S? Need to make sure there are not two correct answers (0 & B) 92 H S? May be RO Knowledge. Based on the information provided could an RO not answer the question once it is concluded that RCS leakage is occurring. You could leave on the procedure number and selected correct answer based on name of procedure only Excessive Leakage. 93 F 1 U LOO. Identify a system where oxygen limits is required to be maintained to prevent corrossion.

1. 2. 3. Psychometric Flaws 4. Job Content Flaws 5. Other 6. 7. Q# LOK LOD I (F/H) (1-5) Stem Cues T/F Credo Partial Job-Minutia #1 Back-Q= SRO U/E/S Explanation Focus Dist. Link units ward KIA Only 94 F S? Could C be correct? 95 F? X U? Not SRO only. Who makes the entry? 96 F 1 U LOD. No knowledge of procedures is required.

Distractors B not plausible. Distractor A could be correct. 97 F 1 X ? U Distractor C & D are not plausible. 98 H 3 S 99 F 1 LOD Common knowledge, RO would know this, but it is the SRO responsibility to know what to do given the conditions. Memory question.... Which AOPs are considered concurrent AOPs? Which AOPs should be performed concurrently while performing procedures in the EOP network? 100 F 1 X X E Could distractor C be correct? Is knowledge of strategy of actions mean to describe the bases? HLC-08 NRC Written Exam 67. Given the following: -The plant is operating at 100% RTP. -A failure of PT-444 input to PC-444J, PZR PRESSURE controller has resulted in an actual Pressurizer pressure increase to 2273 PSIG. -PC-444J has been placed in MANUAL. Which ONE (1) of the following describes the action required to return pressure to 2235 PSIG? A. Decrease the controller output. B. Increase the controller output. C. Lower the pressure setpoint potentiometer adjustment. D. Raise the pressure setpoint potentiometer adjustment. 67 ,,---HLC-08 NRC Written Exam 67.02.1.28 OOl/CONDUCT OF OPERATION/3/4.l/4.1IROIHIOH/N/AIRNP AUDIT -2001IPZR-006 Given the following: -The plant is operating at 100% RTP. -A failure of PT-444 input to PC-444J, PZR PRESSURE controller has resulted in an actual Pressurizer pressure increase to 2273 PSIG. -PC-444J has been placed in MANUAL. Which ONE (1) of the following describes the action required to return pressure to 2235 PSIG? A. Decrease the controller output. 8:0'" Increase the controller output. C. Lower the pressure setpoint potentiometer adjustment. D. Raise the pressure setpoint potentiometer adjustment. The correct answer is B. A: Incorrect -Decreasing the controller output would cause PC-444J to increase pressure. B: Correct -Required to lower pressure, same as pressure deviation for spray valve operation, except it must be done manually. In auto, the input would have to be 25 -50 PSIG above the setpoint to induce spray. In manual, the input is manually lowered to energize heaters, or raised to de-energize heaters or open spray valves. C: Incorrect -With PC-444J in MANUAL, adjustment of the POT would NOT have any effect on PC-444J operation. D: Incorrect -With PC-444J in MANUAL, adjustment of the POT would NOT have any effect on PC-444J operation. Exam Question Number: 67

Reference:

SD-059, PZR, Pages 17 and 18. KA Statement: Knowledge of the purpose and function of major system components and controls. History: Direct from Bank. 74 SD-059 PRESSURIZER SYSTEM 5.1.1 PZR Pressure Control (PZR-Figure 6 & PZR-Figure

7) Pressure control is accomplished via pressure controller PC-444A which is a Proportional

+ Integral controller; the Derivative section has been defeated. This means the controller develops an output signal that is determined by how far pressure is from setpoint (Proportional) and how long the pressure has been away from setpoint (Integral). PT-444 sends a pressure signal to PC-444A which is compared to the pressure setpoint developed by PC-444J which is controlled on the RTGB. PC-444J is a Hagan Auto station with a 10 turn pot capable of developing a control setpoint over the entire pressure range of PT-444. PT-444 ranges from 2500 to 1700 psig therefore PC-444J is capable of 800 psi range of control. For Example if the operator desires the controller to maintain normal pressure of 2235 psig the pot setting would be determined as follows: 2235 -1700

• 10 = 6.69 on the 10 turn pot. 800 The output of PC-444J (setpoint signal) is sent to PC-444A to be compared to the actual pressure.

PC-444A has a gain of 2 which effectively cuts in half the range of control of PZR pressure to 400 psi around the setpoint determined by PC-444J. The controller output is then directed to the proportional heaters, spray valves via controllers PC-444C and PC-444D, backup heaters, PZR PORV 456 and PI-458 and is displayed on the meter on PC-444J The components operated by PC-444A operate at a fixed deviation from setpoint or controller output as observed on the meter on PC-444J, no matter what setpoint is dialed in on PC-444J. For example the backup heaters are set to turn on 25 psi below set pressure. If set pressure is 2235 psig, their setpoint would be 2210 psig and the control output when they came on would be as follows: 2210-2035 400 = .4375 or 43.75% demand If the pot on PC-444J were then set at 6.25 this would give a set pressure of 2200 psig. When the output of PC-444A was at 43.75% the backup heaters would come on, pressure would be 2175 psig; 25 psi below set pressure. The setpoints normally listed for heater, spray, and PCV -456 setpoints are based on a set pressure of 2235 psig where PC-444J is normally set. As stated before, PC-444A is a Proportional + Integral controller, therefore controller output may not correspond exactly to the pressure monitored by the operator. If pressure is away from setpoint for an extended period of time the controller output may saturate while increasing its output trying to return pressure to setpoint. PZR Page 17 of 27 Revision 9 INFORMATION USE ONLY SD-059 5.1.2 PZR Pressure Control Setpoint (PZR-Figure

8) 1. 2. PZR Pressure Controller (PC-444A)

Proportional gain Reset time constant Rate time constant Pressure set point, Pref Spray Valve Controllers (PC-444C, PC-444D) Proportional gain in % spray valve Lift per psi Set point where spray is initiated on compensated pressure signal from PC-444A Setpoint where spray is full open 3. Variable Heater Controller Proportional gain in % heating power Set point where proportional heating 4. 5. 6. is always full on, on signal from PC-444A Setpoint where proportional heating is always full off Power Relief Valve, PCV -455C operating on compensated pressure signal from PC-444A to PC-444B Back-up heaters turned on, due to compensated pressure signal from PC-444A to PC-444F Back-up heaters turned off Power Relief Valve (PCV -456) actuated from actual pressure (PC-445A) 5.1.3 PZR PORV Control (PZR-Figure 7 & PZR-Figure

11) PRESSURIZER SYSTEM 2 12 sec off 2235 psig 2%/psi + 25 psi (2260 psig) +75 psi (2310 psig) -3.33%/psi

-15 psi (2220 psig) + 15 psi (2250 psig) + 100 psi (2335 psig) -25 psi (2210 psig) -15 psi (2220 psig) 2335 psig The PZR PORVs have two modes of control, Normal and Low Temperature Overpressure Protection (LTOPP). In normal mode the PORVs have a permissive of 2000 psig to open PZR Page 18 of 27 Revision 9 INFORMATION USE ONLY

1. 027 AAl.03 00111111111 Given the following:

QUESTIONS REPORT for AUDIT (j0)/)

• The plant is operating at 100% power.
• A failure of the PT-444 input to PC-444J, Pressurizer Pressure Master Controller, caused actual pressurizer pressure to increase to 2273 psig.
• PC-444J has been placed in MANUAL. . WHICH ONE (1) of the following describes the action required to return pressure to 2235 psig? A. Decrease the controller output. 8:1 Increase the controller output. C. Lower the pressure setpoint adjustment.

D. Raise the pressure setpoint adjustment. Required to lower pressure is same as pressure deviation of for spray operation, except it must be done manually. In auto, the input would have to be 25 -50 psig above the setpoint to induce spray. In manual, the input is manually lowered to energize heaters, or raised to de-energize heaters or open spray valves. Common Question 045 Tier 1 Group 1 KIA Importance Rating -RO 3.61 SRO 3.6 Ability to operate and 1 or monitor the following as they apply to the Pressurizer Pressure Control Malfunctions: Pressure control when on a steam bubble. Reference(s) -PZR SO Proposed References to be provided to applicants during examination -None Learning Objective -Question Source -Modified Question History -2002 Audit Exam Question Cognitive Level -Comprehension 10 CFR Part 55 Content -41 Comments -Not exactly KA but should suffice, couldn't figure out a way to write a new one for this KA Category 1: Category 2: Category 3: Category 4: Category 5: Category 6: Category 7: Category 8: Tuesday, June 10, 20084:30:28 PM 1 H LC-08 NRC Written Exam 68. Given the following: -The crew has discovered an error in an AOP. The AOP identified the proper title of a pressure indicator, but describes the indicator as PI-2098 instead of the correct number PI-2089. -A temporary procedure change has been initiated to correct the error. Which ONE (1) of the following statements applies to the temporary procedure change? A. The expiration date of the temporary change is 21 days from the approval date. B. Since the change is to an AOP, it must be considered an INTENT change. C. The temporary change must be approved by the procedure owner (Operations Manager) prior to use. D. The expiration date of the temporary change is 4 months from the approval date. 68 HLC-08 NRC Written Exam 68. G2.2.6 OOllEQUIPMENT CONTROLl3/3.0IRO/LOWININNEW -2008IPRO-NGGC-0204 Given the following: -The crew has discovered an error in an AOP. The AOP identified the proper title of a pressure indicator, but describes the indicator as PI-2098 instead of the correct number PI-2089. -A temporary procedure change has been initiated to correct the error. Which ONE (1) of the following statements applies to the temporary procedure change? A'I The expiration date of the temporary change is 21 days from the approval date. B. Since the change is to an AOP, it must be considered an INTENT change. C. The temporary change must be approved by the procedure owner (Operations Manager) prior to use. D. The expiration date of the temporary change is 4 months from the approval date. The correct answer is A. A: Correct -RNP requirement is that a temporary Change will expire in 21 days. B: Incorrect -A change of intent can NOT be processed as a temporary change. Whether the change is to an AOP has NO bearing on whether the change can be processed as a temporary change. C: Incorrect -A temporary change has to be approved by Management personnel, but it does NOT require the Procedure Owner to approve it before issue. D: Incorrect -Four months is the expiration date for ALL other Progress Energy sites. Exam Question Number: 68

Reference:

PRO-NGGC-0204, Page 32. KA Statement: Knowledge of the process for making changes to procedures. History: New -Written for HLC-08 NRC Exam. 75 9.3 Temporary Change Process NOTE: If possible, the procedure Sponsor or Approval Authority should be consulted to validate the appropriateness of processing a Temporary Change versus a normal procedure revision.

1. Ensure Change meets the definition of a temporary change. 2. Obtain copy of current procedure's title page and affected pages. 3. Increase the revision level to the next alpha designator on the title page and affected pages by using the Passport Controlled Document module to verify the next alpha designator.

4. Mark-up the desired changes. 5. Describe the changes and the reason(s) for the changes on the Temporary Change Form, Attachment

6. 6. Enter the expiration date of the change on Temporary Change Form, Attachment

6. The expiration date will not exceed four months from the ______ interim approval date

[BNP, CR3, HNP]; [not to exceed 21 days from the approval date at RNP]. [R9] 7. Record signatures in designated fields of the Temporary Change Form, Attachment

6. 8. IF the procedure is an NGGC, THEN ensure the other sites evaluate applicability at their facility.

I PRO-NGGC-0204 Rev. 12 32 of 66/ HLC-08 NRC Written Exam 69. Given the following: -The plant is operating at 100% RTP. -All control systems are in automatic. -All required plant equipment is in service. Which ONE (1) of the following equipment failures requires the EARLIEST ACTION per Technical Specifications? Failure of ... A. EDG "A". B. Battery Charger "A". C. Inverter "A". D. DC Bus "A". 69 HLC-08 NRC Written Exam 69.02.2.40 OOllEQUIPMENT CONTROL/3/3.4/4.7JRO/LOW/N/A/NEW -200S/EDO-009 Given the following: -The plant is operating at 100% RTP. -All control systems are in automatic. -All required plant equipment is in service. Which ONE (1) of the following equipment failures requires the EARLIEST ACTION per Technical Specifications? Failure of ... A'I EDG "A". B. Battery Charger "A". C. Inverter "A". D. DC Bus "A". The correct answer is A. A: Correct -Action required within 1 hour to verify breaker alignment for off-site power circuit per LCO 3.S.1, Condition B. B: Incorrect -Battery charger failure must be corrected within 2 hours lAW LCO 3.S.4, Condition A. C: Incorrect -Failure of Instrument Bus 2 requires action within S hours lAW LCO 3.S.7, Condition A. D: Incorrect -Failure of DC Bus A requires action within 2 hours lAW LCO 3.S.4, Condition A. Exam Question Number: 69

Reference:

ITS 3.S.1, 3.S.4 and 3.S.7. KA Statement: Ability to apply Technical Specifications for a system. History: New -Written for HLC-OS NRC Exam 76 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources--Operating AC Sources--Operating 3.8.1 LCO 3.8.1 The following AC electrical sources shall be OPERABLE:

a. The qualified circuit between the offsite transmission network and the onsite emergency AC Electrical Power Distribution System; and Two diesel generators (DGs) capable of supplying the onsite emergency power distribution subsystem(s).

APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS ------------------------------------- NOTE ------------------------------------ LCO 3.0.4.b is not applicable to DGs. CONDITION A. The qualified offsite circuit inoperable. HBRSEP Unit No. 2 A.l AND A.2 REQUIRED ACTION Declare required feature(s) with no offsite power available inoperable when its redundant required feature(s) is inoperable. Restore offsite circuit to OPERABLE status. 3.8-1 COMPLETION TIME 12 hours from discovery of no offsite power to one train concurrent with inoperability of redundant required feature(s) . 24 hours AND 8 days from discovery of failure to meet LCO (continued) Amendment No. 203 ---ACTIONS (continued) CONDITION

---

==> B. One DG i noperab 1 e . -HBRSEP Unit No. 2 REQUIRED ACTION B.1 Perform SR 3.8.1.1 for the offsite circuit. AND B.2 Declare required AC Sources -Operat i n9 3.B.1 COMPLETION TIME ( -----.. ---..

AND Once per 12 hours thereafter 4 hours from feature(s) supported by the inoperable DG discovery of inoperable when its required redundant feature(s) is inoperable. AND B.3.1 Perform SR 3.8.1.2 for OPERABLE DG OR Condition B concurrent with inoperability of redundant required feature(s) 24 hours B.3.2.1 Determine OPERABLE DG 24 hours is not inoperable due to common cause failure. AND B.3.2.2 Perform SR 3.8.1.2 for OPERABLE DG. 3.8*2 96 hours (continued) Amendment No. 176 ---3.8 ELECTRICAL POWER SYSTEMS 3.8.4 DC Sources -Operati n9 DC Sources -Operati ng 3.8.4 LCO 3.8.4 The Train A and Train B DC electrical power subsystems shall be OPERABLE. APPLICABILITY: MODES I. 2. 3. and 4. ACTIONS CONDITION REQUIRED ACTION A. One DC electrical A.1 Restore DC electrical ( power subsystem inoperable. power subsystem to OPERABLE status. B. Required Action and B.1 Be in MODE 3. Associated Completion Time not met. AND B.2 Be in MODE 5. SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.8.4.1 Verify battery terminal voltage is 125.7 V on float charge. HBRSEP Unit No. 2 3.8-19 COMPLETION TIME 6 hours 36 hours FREQUENCY 7 days (continued) Amendment No. 176 ---AC Instrument Bus Sources -Operating 3.8.7 3.8 ELECTRICAL POWER SYSTEMS 3.8.7 AC Instrument Bus Sources-Operating LCO 3.8.7 The following AC Instrument Bus Power Sources shall be OPERABLE:

a. Inverters A and B, and b. Constant Voltage Transformers (CVT) 1 and 2. APPLICABILITY:

MODES 1, 2, 3, and 4. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One AC Instrument Bus A.1 *********NOTE******** power source Enter applicable inoperable. Conditions and Required Actions of LCO 3.8.9, "Distribution Systems . Operating" with any instrument bus de*energized . ... -... _ ............. Restore AC Instrument Bus Power Source to OPERABLE status. B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion AND Time not met. B.2 Be in MODE 5. 36 hours HBRSEP Unit No. 2 Amendment No. 176 HLC-08 NRC Written Exam 70. Which ONE (1) of the following is the radiation dose limit for a DECLARED PREGNANT Robinson employee? A. 50 mRem during the entire pregnancy. B. 500 mRem during the entire pregnancy. C. 200 mRem/year. D. 2000 mRem/year. 70 HLC-08 NRC Written Exam 70. G2.3.4 OOllRADIATION CONTROU3/3.2IROILOW/N/A/NEW -200S/IOCFR20 Which ONE (1) of the following is the radiation dose limit for a DECLARED PREGNANT Robinson employee? A. 50 mRem during the entire pregnancy. 500 mRem during the entire pregnancy. C. 200 mRem/year. D. 2000 mRem/year. The correct answer is B. A: Incorrect -Annual dose limit for the public. B: Correct -10CFR20 limit for a declared pregnancy. C: Incorrect -Annual limit for minors. D: Incorrect -Robinson administrative limit for employees. Exam Question Number: 70

Reference:

1 OCFR20.1208; DOS-NGGC-0002, Pages 32 amd 35. KA Statement: Knowledge of radiation exposure limits under normal or emergency conditions. History: New -Written for HLC-08 NRC Exam. 77 10 CFR 20.1208 Dose equivalent to an embryo/fetus. Page 1 of 1 Index I Site Map I FAQ I Facility Info I Reading Rm I New I Help I Glossary I Contact Us . Google Custom Search: ** Search Options Home> Electronic Reading Room> Document Collections> NRC Regulations (10 CFR) > Part Index> § 20.1208 Dose to an embryo/fetus. § 20.1208 Dose equivalent to an embryo/fetus. (a) The licensee shall ensure that the dose equivalent to the embryo/fetus during the entire pregnancy, due to the occupational exposure of a declared pregnant woman, does not exceed 0.5 rem (5 mSv). (For recordkeeping requirements, see § 20.2106.) (b) The licensee shall make efforts to avoid substantial variation above a uniform monthly exposure rate to a declared pregnant woman so as to satisfy the limit in paragraph (a) of this section. (c) The dose equivalent to the embryo/fetus is the sum of--(1) The deep-dose equivalent to the declared pregnant woman; and (2) The dose equivalent to the embryo/fetus resulting from radionuclides in the embryo/fetus and radionuclides in the declared pregnant woman. (d) If the dose equivalent to the embryo/fetus is found to have exceeded 0.5 rem (5 mSv), or is within 0.05 rem (0.5 mSv) of this dose, by the time the woman declares the pregnancy to the licensee, the licensee shall be deemed to be in compliance with paragraph (a) of this section if the additional dose equivalent to the embryo/fetus does not exceed 0.05 rem (0.5 mSv) during the remainder of the pregnancy. [56 FR 23396, May 21, 1991, as amended at 63 FR 39482, July 23, 1998] Privacy Policy I Site Disclaimer Tuesday, May 27,2008 ATTACHMENT 6 Sheet 1 of 1 Declaration of Pregnancy I hereby do voluntarily declare to Progress Energy that I am pregnant. My best estimate of the date of conception is: ____ (mm/dd/yy). I understand that the radiation dose to my embryo/fetus during my entire pregnancy will not be allowed to exceed 500 mrem. I also understand that, in order to meet the lower dose limit, a change in job or job responsibilities may be required. I understand that I may revoke this declaration at any time by providing written notification to Progress Energy supervision. I agree to abide by all restrictions deemed necessary by Progress Energy to keep the occupational radiation exposure to my unborn child below the limits of 10 CFR 20.1208. Name (Print): SSN: ______________ _ Signature: Date: _______ __ Acknowledged: Date: _______ _ Progress Energy Supervision To Be Completed By Dosimetry Personnel The non-Progress Energy dose to the embryo/fetus from conception to declaration: (Form DOS-NGGC-0002-6-16) Vital Record *1 DOS-NGGC-0002 mrem Rev. 24 Page 32 of 50 I ATTACHMENT 9 Sheet 1 of 2 Information on Embryo/Fetus Exposure Monitoring

1. The NRC has established a limit of SOO mrem to the embryo/fetus during the entire pregnancy.

This limit applies only when the woman declares the pregnancy to Progress Energy in writing. Prior to that declaration Progress Energy has no legal obligation or right to restrict the woman's work assignments for the purpose of controlling the risk of radiation exposure to the embryo/fetus. Each pregnant woman is solely responsible for deciding to declare her pregnancy in writing to Progress Energy supervision. . 2. Progress Energy supervision shall ensure the occupational radiation exposure of each declared pregnant woman is kept below applicable NRC dose limits. A declared pregnant woman may be temporarily reassigned to a different job with equal pay to keep radiation exposure during the period of pregnancy within NRC dose limits. 3. The appropriate time frame for dose calculation is the duration of the pregnancy. It is not appropriate to use a SO year committed dose equivalent or committed effective dose equivalent. The dose from internally deposited radionuclides includes body burdens existing at the time of conception as well as those resulting from subsequent intakes. A threshold value of 1 percent of the stochastic ALI may be used in determining whether a body burden existing at the time of conception should be taken into account in determining the dose to the embryo/fetus.

4. Dosimetry personnel will determine the external dose to the embryo/fetus from the conception date to the declaration date based on the deep dose to the mother. If the conception date falls in the middle of an exposure period, subtract any secondary dosimeter results for the period before the conception date from the primary dosimeter results for that period in determining the amount of dose received so far during the pregnancy.

In the absence of secondary dosimeter data, assume the dose was received at a uniform rate throughout the exposure period during which the conception date falls and prorate the embryo/fetus dose based on the deep dose equivalent for the declared pregnant woman. S. If the dose since conception exceeds 450 mrem, then set the available dose to 50 mrem for the remainder of the pregnancy. If the dose since conception is less than or equal to 450 mrem, then set the available dose equal to the difference between 500 mrem and the dose since conception.

6. A declared pregnant woman may revoke her declaration of pregnancy at any time. After revocation, her administrative dose limits will be reset and any further dose to the embryo/fetus shall not be recorded.

I DOS-NGGC-0002 Rev. 24 Page 35 of 50 I HLC-08 NRC Written Exam 71. Given the following: -You are in the Spent Fuel Pit performing fuel handling operations. -A fuel handling accident occurs, and an evacuation is ordered via a PA announcement. Which ONE (1) of the following is the area designated as the assembly point for this event? A. Outside the Spent Fuel Pit Building door. B. Landing halfway down the stairs outside the Spent Fuel Pit Building. C. Outside the New Fuel Handling Building door. D. North side of the RWST area. 71 HLC-08 NRC Written Exam 71. G2.3.13 OOllRADIATION CONTROLl3/3.4/3.8IRO/LOW/NINNEW -2008/AOP-013-004 Given the following: -You are in the Spent Fuel Pit performing fuel handling operations. -A fuel handling accident occurs, and an evacuation is ordered via a PA announcement. Which ONE (1) of the following is the area designated as the assembly point for this event? A. Outside the Spent Fuel Pit Building door. Landing halfway down the stairs outside the Spent Fuel Pit Building. C. Outside the New Fuel Handling Building door. D. North side of the RWST area. The correct answer is B. A: Incorrect -SFP personnel are expected to evacuate the SFP area, but transit to the landing halfway down to ensure shielding is provided. (Distance) B: Correct -AOP-013 specifies SFP personnel to stop at the landing halfway down the stairs (shielded, but prevents spread of contamination until cleared.) C: Incorrect -This is the assembly point for an accident in the New Fuel Handling Building. D: Incorrect -North side of the RWST would allow unecessary spread of contamination. Exam Question Number: 71

Reference:

AOP-013, Page 4. KA Statement: 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. History: New -Written for HLC-08 NRC Exam. 78 /"" Rev. 11 AOP-013 FUEL HANDLING ACCIDENT INSTRUCTIONS

1. Evacuate Personnel From Affected Area: a. Place the VLC switch in EMERG b. Check accident location -INSIDE CV c. Depress and hold CV EVACUATION .HORN button for 15 seconds d. Announce the following over the PA system: 1) nature of the accident 2) location of the accident 3) location of personnel assembly area based on accident location:

• CV -outside airlock at bottom of stairs Spent Fuel pit -landing halfway down the stairs
• New Fuel Storage Area -outside Fuel Handling Building door
• Spent Fuel Shipping Cask -Evacuate Area Only (assembly area based on location)

e. Repeat the alarm/horn for 15 seconds f. Repeat the announcement Page 4 of RESPONSE NOT OBTAINED b. Perform the following:

1) Place and hold EVACUATION ALARM switch in the LOCAL position for 15 seconds. 2) Go To Step 1.d. 11 HLC-08 NRC Written Exam 72. Given the following:

-While entering the plant to relieve the shift, you notice smoke coming from the Auxiliary Building. -You have NOT logged onto any RWP or obtained dosimetry. -You feel that your assistance in the Auxiliary Building is needed. Which ONE (1) of the following is allowed/required lAW the plant Radiation Protection Standards for the stated conditions? A. An exception to the requirement for proper dosimetry AND RWP login is allowed for qualified members of the Fire Brigade. B. Proper dosimetry AND RWP login must be completed prior to Auxiliary Building entry. NO exceptions are allowed for the stated circumstances. C. An exception for the requirement for proper dosimetry AND RWP login is allowed IF an RC Tech is present to monitor personnel entry/exit to the Auxiliary Building. D. RWP login is NOT required, and emergency dosimetry is obtained from the Work Control Center. 72 HLC-08 NRC Written Exam 72. G2.3.7 OOllRADIATION CONTROU3/3.5/3.6IROILOW/N/A/NEW -20081 Given the following: -While entering the plant to relieve the shift, you notice smoke coming from the Auxiliary Building. -You have NOT logged onto any RWP or obtained dosimetry. -You feel that your assistance in the Auxiliary Building is needed. Which ONE (1) of the following is allowed/required lAW the plant Radiation Protection Standards for the stated conditions? A. An exception to the requirement for proper dosimetry AND RWP login is allowed for qualified members of the Fire Brigade. Proper dosimetry AND RWP login must be completed prior to Auxiliary Building entry. NO exceptions are allowed for the stated circumstances. C. An exception for the requirement for proper dosimetry AND RWP login is allowed IF an RC Tech is present to monitor personnel entry/exit to the Auxiliary Building. D. RWP login is NOT required, and emergency dosimetry is obtained from the Work Control Center. The correct answer is B. A: Incorrect -Exception for dosimetry and RWP login is NOT allowed. Operations pesonnel are required to obtain proper dosimetry prior to relieving their watchstations. B: Correct -RWP login and proper dosimetry is required prior to entry into any RCA. C: Incorrect -RC Tech being present does NOT relieve personnel of the requirement for proper dosimetry. D: Incorrect -Emergency dosimetry is NOT available in the Work Control Center. Exam Question Number: 72

Reference:

OMM-001-12, Page S. KA Statement: Ability to comply with radiation work permit requirements during normal or abnormal conditions. History: New -Written for HLC-OS NRC Exam. 79 INFORMATION USE 8.0 INSTRUCTIONS{tc "INSTRUCTIONS" \f C \I I} 8.1 General Shift Relief Guidelines{tc "General Shift Relief Guidelines" \f C \I2} 8.1.1 All shift turnovers will be performed at the normal watchstation for the personnel being relieved. During abnormal conditions shift personnel may be relieved *at places other than normal watchstations (example: during the performance of OSTs). 8.1.2 The turnover sheets will be used and referred to during the performance of shift turnover. 8.1.3 Turnover during the approach to criticality should be avoided. Major evolutions in progress should be completed prior to relief of on-shift personnel. Evolutions in progress will be turned over only with the approval of the SSO. (SOER 07-1, Recommendation

2) ---?13.1,4 All Operations Shift Personnel will obtain dosimetry prior to relieving the watch. (NCR 81852, CR 99-00807) 8.1.5 Personnel who require corrective lenses are to maintain respirator glasses in a specified storage location lAW OMM-001-5.

Personnel wearing gas permeable contact lenses are not required to verify respirator glasses for Control Room watches or Fire Brigade assignment. 8.1.6 Personnel who will be on the Fire Brigade are to verify proper storage and completeness of their fire turnout gear. IOMM-001-12 Rev. 56 Page 8 of 661 HLC-08 NRC Written Exam 73. Given the following: -The crew is responding to a RED PATH on the Heat Sink CSFST. -The STA has just reported a RED PATH on the Integrity CSFST. FRP-H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK. FRP-P.1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITION. Which ONE (1) of the following is the crew's response to the RED PATH on Integrity? A. Continue with FRP-H.1; Heat Sink CSF has a higher priority than the Integrity CSF. B. Transition to FRP-P.1; Integrity CSF has a higher priority than the Heat Sink CSF. C. CRSS must implement FRP-H.1 OR FRP-P.1 depending on which one will most effectively mitigate the current plant conditions. D. Continue with FRP-H.1, while concurrently implementing FRP-P.1 to ensure that ALL RED path FRPs are addressed. 73 .. ' HLC-08 NRC Written Exam 73. G2.4.16 OOllEMERG PROCIPLAN/3/3.5IROILOW/NINNEW -2008/0MM-022-003 Given the following: -The crew is responding to a RED PATH on the Heat Sink CSFST. -The STA has just reported a RED PATH on the Integrity CSFST. FRP-H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK. FRP-P.1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITION. Which ONE (1) of the following is the crew's response to the RED PATH on Integrity? A'! Continue with FRP-H.1; Heat Sink CSF has a higher priority than the Integrity CSF. 8. Transition to FRP-P.1; Integrity CSF has a higher priority than the Heat Sink CSF. C. CRSS must implement FRP-H.1 OR FRP-P.1 depending on which one will most effectively mitigate the current plant conditions. D. Continue with FRP-H.1, while concurrently implementing FRP-P.1 to ensure that ALL RED path FRPs are addressed. The correct answer is A. A: Correct -Heat Sink has a higher priority. B: Incorrect -Heat Sink is a higher priority and will be continued. C: Incorrect -Implementation of RED PATH FRPs are NOT discretionary based on the CRSS and plant conditions. If a higher CSF RED path is received, OMM direction is to Go To that CSF. D: Incorrect -Only one FRP is to be implemented at a time, highest priority takes precedence. Exam Question Number: 73

Reference:

OMM-022, Pages 18, 19. KA Statement: 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. History: New -Written for HLC-08 NRC exam. 80 8.2.6 (Continued) NOTE: WHITE is NOT a color coding for a terminus. This is a function of ERFIS only to indicate that ERFIS does not have sufficient input to make the determination of the proper terminus. IOMM-022 7. Color-coding can be either RED, ORANGE, YELLOW, or GREEN, with GREEN representing a "SAT" safety status. Each non-green color represents an action level that should be addressed according to the Rules of Priority for Status Tree Use. 8. Several special conditions also affect the CSFSTs indicated by ERFIS:

• All CSFSTs are forced to a GREEN-condition when the plant mode is Cold Shutdown.
• The Heat Sink Tree is forced to a GREEN-condition when the plant is less than 350°F.
• The Subcriticality Tree is forced to a GREEN-condition when the plant mode is Power Operation or Hot Shutdown except:
• When in the Power Operation mode, the actual Critical Safety Function Status will be displayed if the Reactor Trip and Bypass Breakers are open.
• When in any mode, the actual Critical Safety Function Status will be displayed if a trip condition exists (as determined from Reactor Protection System inputs). 9. The six Status Trees are always evaluated in the following sequence (order of priority):

.-1) Subcriticality (S) 2) Core Cooling (C) 3) Heat Sink (H) 4) Integrity (P) 5) Containment (J) 6) Inventory (I) Rev. 28 Page 18 of 541 8.2.6 (Continued) IOMM-022 10. If identical color priorities are found on different trees during monitoring, the required action priority is determined by the above sequence. For example, a RED-condition on the Subcriticality Tree takes priority over a RED-condition on Core Cooling Tree. 11. The user begins monitoring with the Subcriticality Tree. Questions are answered based on plant conditions at the time, and the appropriate branch line followed to the next question. An individual Status Tree evaluation is complete when the user arrives at a color-coded terminus. With the exceptions noted below, the color and instructions of the terminus are noted and the user continues to the next tree in sequence.

a. c. d. e. If any RED terminus is encountered, the operator is required to immediately stop any Path or EPP in progress, and to perform the Function Restoration Procedure (FRP) required by the terminus.

If, during the performance of any RED-condition FRP, a RED-condition of higher priority arises, then the higher priority condition should be addressed first, and the lower priority RED-condition FRP suspended. If any ORANGE terminus is encountered, the operator is expected to monitor all of the remaining trees and if no RED-condition is encountered, suspend any Path or EPP in progress and perform the FRP required by the ORANGE terminus. If during the performance of an ORANGE-condition FRP, any RED-condition or higher priority ORANGE-condition arises, then the RED or higher priority ORANGE-condition is to be addressed first, and the original condition FRP suspended. An exception to item 8.2.6.11.d above is that if an ORANGE condition goes to RED AND the same FRP is applicable, then progress in the FRP should continue without starting over. Rev. 28 Page 19 of 541 HLC-08 NRC Written Exam 74. Which ONE (1) of the following describes the reason for the sequence of diagnostic steps in PATH-1? Main Steam Line Break diagnosis takes priority because ... A. indications could potentially mask other failures. B. early isolation of AFW is required to maintain CV within accident analysis. C. of the high potential for personnel injury during Steam Break accidents outside of CV. D. of the potential for CV failure if all S/Gs are faulted. 74 HLC-08 NRC Written Exam 74. G2.4.22 OOllEMERG PROC/PLAN/3/3.6/4.4IROILOW/N/A/NEW -200S/PATH-I-007 Which ONE (1) of the following describes the reason for the sequence of diagnostic steps in PATH-1? Main Steam Line Break diagnosis takes priority because ... A:' indications could potentially mask other failures. B. early isolation of AFW is required to maintain CV within accident analysis. C. of the high potential for personnel injury during Steam Break accidents outside of CV. D. of the potential for CV failure if all SIGs are faulted. The correct answer is A. A: Correct -A Main Steam break causes a rapid RCS cooldown and SI until the SIG blows dry. RCS temperature will stabilize and the RCS will repressurize to restore normal RCS conditions. This allows other accidents to be diagnosed or the EOP network exited if conditions allow. B: Incorrect -Early isolation of AFW will limit RCS cooldown and CV pressure, but it is NOT required to keep CV pressure within analysis. C: Incorrect -Potential for personnel injury is limited to area adjacent to CV wall. Any steam break downstream of the MSIVs will be terminated by automatic closure of the MSIVs. D: Incorrect -EPP-16 is the procedure that will combat ALL 3 faulted SIGs and bounds the failure within CV analysis. Exam Question Number: 74

Reference:

PATH-1 SO, Page 3. KA Statement: Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations. History: New -Written for HLC-08 NRC exam, 81 DISCUSSION (From the WOG E-O Basis Document)

1. INTRODUCTION Guideline E-O, REACTOR TRIP OR SAFETY INJECTION, provides actions to verify proper response of the automatic protection systems following manual or automatic actuation of a reactor trip or safety injection, to assess plant conditions, and to identify the appropriate Optimal Recovery Guideline.

Guideline E-O is to be entered when any of the following occur: 1) A reactor trip is required as determined by plant specific setpoints or requirements being exceeded.

2) A reactor trip has occurred as determined by the plant annunciators, neutron flux instrumentation, and control rod position indicators.

3) A safety injection is required as determined by plant specific setpoints or requirements being exceeded.

4) A safety injection has occurred as determined by the plant annunciators, SI pump status, or other plant specific means. Once E-O is entered, it is not exited until there is a direct transition to an Optimal Recovery Guideline (ORG) as directed by the symptoms being monitored in E-O or to a Function Restoration Guideline (FRG) as directed by the Critical Safety Function Status Trees or symptoms being monitored in E-O. 2. DESCRIPTION Guideline E-O, REACTOR TRIP OR SAFETY INJECTION, provides the operator with the necessary guidance to verify that all automatic actions have occurred as designed and presents the diagnostic sequence to be followed in the identification of the appropriate Optimal Recovery Guideline.

These include: 1. ECA-O.O, LOSS OF ALL AC POWER 2. ES-O.1, REACTOR TRIP RESPONSE 3. E-1, LOSS OF REACTOR OR SECONDARY COOLANT 4. E-2, FAULTED STEAM GENERATOR ISOLATION

5. E-3, STEAM GENERATOR TUBE RUPTURE 6. ES-1.1, SI TERMINATION

7. ECA-1.2, LOCA OUTSIDE CONTAINMENT It is expected that the operator will attempt to take manual actions to correct for anomalous conditions during power operation.

Such actions would include taking manual control of the automatic control systems, turning on additional charging pumps, reducing power level, etc. If these types of actions do not alleviate the trend toward a reactor trip or safety injection, the operator is permitted to trip the reactor and, if necessary, actuate safety injection. The reactor protection equipment is designed to safely shut down the reactor in the event that the anomalous condition cannot be corrected. The safety injection system is designed to provide emergency core cooling water and boration to maintain a safe reactor shutdown condition. The plant safeguards systems operate with offsite electrical power or from onsite emergency diesel-electric power, should offsite power not be available. The operator will enter E-O on a reactor trip or safety injection, whether the signal was automatic or a result of manual actuation. Through symptom-based diagnosis, the operator is directed to the proper Optimal Recovery Guideline to facilitate optimal recovery. Transient descriptions are provided in the appropriate background documents. Many parameters behave in a similar manner for loss of reactor coolant, secondary coolant and steam generator tube ruptures. For example, RCS pressure drops for all three cases. The symptoms used to diagnose the three major event categories are those most representative of the fault. A break in the secondary is diagnosed by secondary pressure decreasing in an uncontrolled manner or any steam generator completely depressurized. A primary to secondary break is diagnosed by abnormal secondary radiation. A loss of primary coolant into containment is diagnosed by abnormal containment pressure, sump level, or radiation. Abnormalities in both containment pressure and sump level can also occur for a secondary break in Q9ntainment. For that reason the secondary pressure is -checked before the containment conditions in the dia nostic steps to eliminate a seconda break as bein the cause of the contamment conditions The rediagnostlc capabilities 0 t esubsequent guidelines, suc . as in E-1, L REACTOR OR SECONDARY COOLANT, allow for any misdiagnosis to be corrected and the operator to proceed to the proper guideline in the Emergency Response Guideline (ERG) network. I PATH-1-BD I Rev 18 Page 3 of 95/ HLC-08 NRC Written Exam 75. Given the following: -A Reactor Trip and Safety Injection has occurred due to a LBLOGA. -The RO immediately notices that Train "A" SI and RHR Pumps did NOT start. -Immediate Actions are complete, but have NOT been verified by the GRSS. What actions are required and why? A. Immediately start non-running EGGS pumps; both trains of EGGS is required to ensure core damage does NOT occur. B. Verify Immediate Actions using PATH-1; ensuring the Reactor and Turbine are safely shutdown and safeguards electrical busses energized are assumptions of PATH-1. C. Wait until after the SI Sequencer is complete to manually start the EGGS pumps; manual operation of EGGS components is inhibited until sequence is complete. D. Wait until after the SI has been reset to manually start the EGGS pumps; manual operation of EGGS components is inhibited until SI has been reset. 75 ,--.,. HLC-08 NRC Written Exam 75. G2.4.23 OOIIEMERG PROCIPLAN/3/3.4/4.4IROIHIGHININNEW -2008/0MM-022-008 Given the following: -A Reactor Trip and Safety Injection has occurred due to a LBLOGA. -The RO immediately notices that Train "An SI and RHR Pumps did NOT start. -Immediate Actions are complete, but have NOT been verified by the GRSS. What actions are required and why? A. Immediately start non-running EGGS pumps; both trains of EGGS is required to ensure core damage does NOT occur. B!'" Verify Immediate Actions using PATH-1; ensuring the Reactor and Turbine are safely shutdown and safeguards electrical busses energized are assumptions of PA TH-1. C. Wait until after the SI Sequencer is complete to manually start the EGGS pumps; manual operation of EGGS components is inhibited until sequence is complete. D. Wait until after the SI has been reset to manually start the EGGS pumps; manual operation of EGGS components is inhibited until SI has been reset. The correct answer is B. A: Incorrect -Immediate Actions are performed from memory and verified by procedure to be considered complete. Gore damage is prevented due to minimum safeguards (1 train) equipment operating. B: Correct -Immediate Actions are performed from memory and must be verified by procedure to be considered complete. Any manual actions must wait until after actions have been performed and verified. G: Incorrect -Manual start of SI equipment is NOT inhibited while the Sequencer is running. Starting any of the equipment prior to the completion of the sequence could result in 2 loads starting at the same time and possbily causing the Emergency Bus feeder breaker to trip on overcurrent. D: Incorrect -Manual start of SI equipment is NOT inhibited until SI is reset. Manually starting EGGS equipment is NOT inhibited at any time. Exam Question Number: 75

Reference:

OMM-022, Page 29. KA Statement: Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations. History: New -Written for HLC-08 NRC Exam. 82 8.3.2 Performance of EOP Steps Early{ TC "Performance ofEOP Steps Early" \fC \}"3"} IOMM-022 1. Performing EOP steps early is allowed, but must be done with caution so as not to mask symptoms or defeat the intent of the EOPs being used. 2. The following guidelines must be met in order to perform actions prior to being directed by the EOP Network. All EOP immediate actions must be completed (and verified) prior to taking any early action or non-EOP action.

• The concurrence of the CRSS must be obtained prior to taking the action.
• The action can not defeat the intent of the procedure or the WOG mitigative strategy.
• Personnel are available to perform the action to the extent that performance of the action will not hinder or delay the performance of the required actions. 3. The RTGB Operators should not overload the CRSS with requests for preemptive actions that would distract from the precise and timely completion of the procedures.

The CRSS must use command and control if requests for preemptive actions are distracting from the completion of the procedure.

4. The following examples either demonstrate the advantages of or highlight the need for caution when performing EOP steps out-of-sequence:

Advantage:

• Isolation of steam to the SDAFW Pump during a SGTR prior to it starting will prevent a potential release of radioactive effluent.
• Control of AFW flow to a known ruptured S/G to between 8% and 50% will enhance the ability to prevent flooding of the S/G. Rev. 29 Page 29 of 54/}}