ML102780356

Initial Exam 2010-301 Draft RO Written Exam

ML102780356
Person / Time
Site:	Surry
Issue date:	09/14/2010
From:	NRC/RGN-II
To:	Virginia Electric & Power Co (VEPCO)
References
50-280/10-301, 50-281/10-301
Download: ML102780356 (127)

Text

1. 000000000007EK3.01 1 Unit 1 Initial Conditions:

- 15% Power.

- An automatic Reactor Trip occurs.

Current conditions:

- Reactor is verified tripped.

- Main Turbine DID NOT automatically trip.

- Operators attempted to manually trip the Main Turbine from the control room.

- All Turbine Stop and Governor valves remain OPEN.

- Generator output breakers are CLOSED.

- Turbine speed is NOT DECREASING.

- It is now 45 seconds after the Reactor Trip.

Based on the current conditions, which one of the following is (1) the NEXT action required by 1-E-0, "REACTOR TRIP OR SAFETY INJECTION," and (2) the reason for this action?

A. (1) Close MSTVs.

(2) Prevent an uncontrolled cooldown of the Reactor Coolant System (RCS).

B. (1) Close MSTVs.

(2) Prevent a Loss of Heat Sink condition from occurring.

C. (1) Open generator output breakers AND place Excitation control switch in OFF.

(2) Prevent motoring the Main Generator.

D. (1) Open generator output breakers AND place Excitation control switch in OFF.

(2) Actuate an alternate Turbine trip signal.

K/A Reactor Trip 007EK3.01 Knowledge of the reasons for the following as they apply to a reactor trip:

Actions contained in EOP for reactor trip.

(CFR 41.5 / 41.10 / 45.6 / 45.13) (RO - 4.0)

K/A Match Analysis This question matches the K/A statement by requiring the applicants to correctly identify the next required action in step 2 (immediate operator actions) of 1-E-0, given an operationally valid scenario, as well as identify the correct reason for this action. The reason for the next action is given in the Westinghouse EOP Background document/detailed step description; however, it is NOT SRO-only because it is dealing with immediate operator actions which are required to be known from memory by all

licensed operators.

Designated as memory because steps were taken right from E-0.

Answer Choice Analysis A. CORRECT. In accordance with 1-E-0, REACTOR TRIP OR SAFETY INJECTION, step [ 2] b) RNO, which states IF Turbine will NOT Trip, THEN close MSTVs. Therefore, the correct NEXT ACTION is to close the MSTVs. The reason for this step is given in the Westinghouse EOP Background document for E-0 as follows:

The turbine is tripped to prevent an uncontrolled cooldown of the RCS due to steam flow that the turbine would require.

B. INCORRECT. In accordance with 1-E-0, REACTOR TRIP OR SAFETY INJECTION, step [ 2] b) RNO, which states IF Turbine will NOT Trip, THEN close MSTVs. Therefore, the correct NEXT ACTION is to close the MSTVs and part (1) of this distractor is correct. However, part (2) is incorrect; the correct reason for this step is given in the Westinghouse EOP Background document for E-0 as follows: The turbine is tripped to prevent an uncontrolled cooldown of the RCS due to steam flow that the turbine would require. Part (2) is plausible because if the turbine remains on line, it will continue to draw steam and lower steam generator levels, which may result in a loss of heat sink.

C. INCORRECT. Part (1) is incorrect, but plausible, because it is another action contained in step [2] of 1-E-0 to be taken after 30 seconds have passed. In accordance with 1-E-0 step [ 2] d) RNO, which states IF Generator Output Breakers do NOT open within 30 seconds, THEN manually open output breakers AND place the EXCITATION control switch in OFF. Part (2) of distractor C. is plausible because it is one of the reasons for taking the action in part (1).

D. INCORRECT. Part (1) is incorrect as detailed in the analysis for distractor C. Part (2) of distractor D. is plausible because tripping the main generator will cause a turbine trip signal, and the goal of step [2] of 1-E-0 is to trip the turbine.

Supporting References

1. Surry Procedure 1-E-0, REACTOR TRIP OR SAFETY INJECTION, rev. 61.

2. Westinghouse E-0 Background Document HP-Rev. 2, dtd 04/30/2005. Especially section 4.1 Detailed Description of Steps, Notes, and Cautions.

3. This question is modified from Braidwood 1 007EK1.03 from 04/01/1996 to apply to the Surry Power Station and to enhance plausibility.

4. This question is also modified from Beaver Valley 1 007EK1.03 from 04/28/1997 to apply to the Surry Power Station and to enhance plausibility.

References Provided to Applicant None.

Answer: A

2. 000000000008AK2.02 1 Initial conditions:

- Unit 1 is at 100% power.

RC-LT-461 selected for Upper Control Channel.

RC-LT-459 selected for Lower Control Channel.

Current conditions:

- 1C-B8 PZR LO PRESS is lit

- 1C-C8 PRZR HI LVL HTRS ON is lit

- PRZR Master pressure controller output is 0%.

- PRZR level controller output is 0%.

- Pressure trends are observed on the following indicators:

RC-PI-455 = 2245 psig and dropping.

RC-PI-456 = 2243 psig and dropping.

RC-PI-457 = 2135 psig and dropping.

RC-PR-1441, Position 1 = 2130 psig and dropping.

- Level trends are observed on the following indicators:

RC-LT-459 = 48% and dropping.

RC-LT-460 = 47% and dropping.

RC-LT-461 = 68% and rising.

Based on the current conditions, which one of the following explains the above event?

A. PRZR Master pressure controller output has failed low.

B. A leak has developed at the instrument tap to the reference leg for level transmitter 1-RC-LT-461.

C. PRZR level controller output has failed low.

D. A leak has developed at the instrument tap to the variable leg for level transmitter 1-RC-LT-461.

K/A Pressurizer Vapor Space Accident Knowledge of the interrelations between the Pressurizer Vapor Space Accident and Sensors and detectors.

(CFR: 41.7/45.7) (RO - 2.7)

K/A Match Analysis The RO applicant is required to recognize how a leak from the Pressurizer vapor space (i.e., reference leg instrument tap) will effect level and pressure indications and control on the pressurizer.

Answer Choice Analysis A. INCORRECT. Plausible because an output failure would cause the controller output to indicate zero and actual pressure would be trending up due to heaters coming on. However, the failure would not explain the contradictory trends in pressurizer level on the available pressurizer level indications.

B. CORRECT. With a leak at the instrument tap to the reference leg of RC-LT-461, water in the reference leg will boil off causing indicated water level to increase and pressure in the reference leg to drop. In addition, pressure transmitters PT-457, PT-445 (input to pressure recorder PR-1441), and PT-444 (input to Pressurizer Master Pressure Controller) all tap off the same reference leg piping that supplies RC-LT-461 so indicated pressure would decrease rapidly on these pressure transmitters. PT-456 and PT-457 would also decrease. As the heaters come on in response to low pressure sensed by the Pressurizer Master Pressure Controller and an increase in sensed level of more than 5% from program level (53.7%).

C. INCORRECT. Plausible because an output failure would cause the controller output to indicate zero and actual level would be trending down due to less charging flow. However, the failure would not explain the trends on pressurizer pressure or the zero output on the Pressurizer Master Controller.

D. INCORRECT. Plausible if the applicant believed a leak on the variable leg caused indicated water level to increase.

Supporting References

1. Surry lesson plan ND-93.3-LP-7, "Pressurizer Level Control System," rev. 10, Obj D, pp. 5-6.

2. Surry lesson plan ND-93.3-LP-5, "Pressurizer Pressure Control System," rev. 14, Obj D, pp. 6-7.

References Provided to Applicant None.

Answer: B

3. 000000000009EK1.01 1 Unit 1 plant conditions:

- SBLOCA has occurred

- RCS hot legs are voided

- FR-C.1 RESPONSE TO INADEQUATE CORE COOLING in progress

- RCPs are off

- Total AFW flow = 100 gpm

- The SRO directs the intact SGs to be depressurized to 200 psig Based on the above conditions, which one of the following: (1) states what limits (if any) are placed on SG depressurization and (2) the reason for this action?

A. (1) Limited such that RCS cooldown rate does not exceed 100 oF/Hr (2) Maintain the SG as a heat sink via reflux boiling B. (1) Limited such that RCS cooldown rate does not exceed 100 oF/Hr (2) To enable the RCPs to be restarted C. (1) There is no limit on how fast you can depressurize the SGs (2) Maintain the SG as a heat sink via reflux boiling D. (1) There is no limit on how fast you can depressurize the SGs (2) To enable the RCPs to be restarted K/A Small Break LOCA. Knowledge of the operational implications of the following concepts as they apply to the small break LOCA: Natural circulation and cooling, including reflux boiling.

K/A Match Analysis Requires applicant to know operational method to depressurize SGs in order to establish reflux boiling during a SBLOCA.

Answer Choice Analysis A. Incorrect: FR-C.1 Step 23 directs SGs to be depressurized at the maximum rate.

1st part is plausible because when not in loss of cooling scenarios, RCS cooldown is limited to prevent PTS concerns. 2nd part is correct.

B. Incorrect: Step 23 directs SGs to be depressurized at the maximum rate. 1st part is plausible because when not in loss of cooling scenarios, RCS cooldown is limited to prevent PTS concerns. 2nd part is plausible because a heat sink needs to be established if forced cooling is to be established.

C. Correct: FR-C.1 Step 23 directs SGs to be depressurized at the maximum rate.

Depressurizing the SGs allows more feedwater flow into the SGs. This will condense some of the steam on the primary side of the U-Tubes which will flow back down the hot leg into the core (Reflux Boiling).

D. Incorrect: 1st part is correct. 2nd part is plausible because a heat sink needs to be established if forced cooling is to be established.

Supporting References ND-95.3-LP-38 Obj: B FR-C.1 RESPONSE TO INADEQUATE CORE COOLING Step 23 References Provided to Applicant none Answer: C

4. 000000000011EG2.4.6 1 Unit 1 Initial Conditions:

- Unit 1 experienced a design-basis Large Break Loss of Coolant Accident (LBLOCA) coincident with a loss of offsite power.

Current Conditions:

- You are arriving at the plant to assume the Operator at the Controls (OATC) position.

- It is now ten (10) hours since the LOCA occurred.

Based upon the current conditions, what is the method of core cooling in use?

A. Recirculation flow from the Containment sump through the SI system with discharge into the cold legs. ONLY one train of ISRS in service.

B. Recirculation flow from the Containment sump through the SI system with discharge into the cold legs. BOTH trains of ISRS in service.

C. Recirculation flow from the Containment sump through the SI system with discharge into the hot legs. ONLY one train of ISRS in service.

D. Recirculation flow from the Containment sump through the SI system with discharge

into the hot legs. BOTH trains of ISRS in service.

K/A Large Break LOCA 011EG2.4.6 Large Break LOCA: Knowledge of EOP mitigation strategies.

(CFR 41.10 / 43.5 / 45.13) (RO - 3.7)

K/A Match Analysis This question matches the K/A statement by requiring the applicant to correctly determine the proper recirculation alignment (EOP mitigation strategy) that would be in use 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after a large-break LOCA occurred, including the ISRS alignment. The applicant needs to recall that a transfer to hot leg recirculation takes place 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after a LBLOCA, and needs to be completed before 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> have elapsed. The Surry alignment closes all discharge valves into the cold legs and aligns all recirculation flow to the hot legs. SURRY Please Verify This Statement: Because of the loss of offsite power, only one train of ISRS will be operating.

Answer Choice Analysis A. INCORRECT. The first part of this distractor is plausible because it is the method of core cooling from the swapover from injection flow to recirculation flow until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> post-LOCA; it should be the most familiar method of recirculation core cooling to the applicant. This choice could be selected if the applicant forgets about the switch to hot leg recirculation, or if the applicant does not remember the correct time frame for swapping to hot leg recirculation. The second part of this distractor is correct; only one train of ISRS will be in service. At step 9 of 1-E-1, SW flow to one train of RS HXs will be secured. Then, at step 12 of 1-E-1, the CS pumps and the operating OSRS pumps are secured. Therefore, at the ten hour point, only one ISRS pump will be operating in recirc providing spray to the header.

B. INCORRECT. The first part of this distractor is plausible because it is the method of core cooling from the swapover from injection flow to recirculation flow until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> post-LOCA; it should be the most familiar method of recirculation core cooling to the applicant. This choice could be selected if the applicant forgets about the switch to hot leg recirculation, or if the applicant does not remember the correct time frame for swapping to hot leg recirculation. The second part of this distractor is incorrect; only one train of ISRS will be in service. At step 9 of 1-E-1, SW flow to one train of RS HXs will be secured. Then, at step 12 of 1-E-1, the CS pumps and the operating OSRS pumps are secured. Therefore, at the ten hour point, only one ISRS pump will be operating in recirc providing spray to the header. Both ISRS pumps is plausible because of the LBLOCA; both will be needed to depressurize containment. Plausibility is further enhanced if the applicant believes that both trains will remain operating based

on both emergency buses being powered from the EDGs.

C. CORRECT. The transfer to hot leg recirculation is required to occur 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> post-LOCA, and be completed after 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> have elapsed. Therefore, the operator arriving at the plant to relieve the watch 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> post-LBLOCA should expect the plant to be on recirculation to the hot legs ONLY. Because no other conditions are given in the stem, the applicant must consider that all systems have functioned as they were designed. The second part of this distractor is also correct; only one train of ISRS will be in service. At step 9 of 1-E-1, SW flow to one train of RS HXs will be secured.

Then, at step 12 of 1-E-1, the CS pumps and the operating OSRS pumps are secured.

Therefore, at the ten hour point, only one ISRS pump will be operating in recirc providing spray to the header.

D. INCORRECT. The transfer to hot leg recirculation is required to occur 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> post-LOCA, and be completed after 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> have elapsed. Therefore, the operator arriving at the plant to relieve the watch 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> post-LBLOCA should expect the plant to be on recirculation to the hot legs ONLY. Because no other conditions are given in the stem, the applicant must consider that all systems have functioned as they were designed. However, the second part of this distractor is incorrect; only one train of ISRS will be in service. At step 9 of 1-E-1, SW flow to one train of RS HXs will be secured. Then, at step 12 of 1-E-1, the CS pumps and the operating OSRS pumps are secured. Therefore, at the ten hour point, only one ISRS pump will be operating in recirc providing spray to the header. Both ISRS pumps is plausible because of the LBLOCA; both will be needed to depressurize containment. Plausibility is further enhanced if the applicant believes that both trains will remain operating based on both emergency buses being powered from the EDGs.

Supporting References

1. Surry Lesson Plan ND-95.2-LP-7, "Loss of Reactor Coolant Accident," rev. 10 dtd 02/13/08. Reflux boiling is described on p. 28.

2. Surry Procedure 1-E-1, "LOSS OF REACTOR OR SECONDAY COOLANT," rev.

31. Note before step 28 states that hot leg recirc must be established by 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> after the event.

3. Surry Lesson Plan ND-95.3-LP-7, "E-1, Loss of Reactor or Secondary Coolant," rev 16 dtd 07.28/08.

4. Surry Procedure 1-ES-1.3, "TRANSFER TO COLD LEG RECIRCULATION," rev.

17.

5. Surry Lesson Plan ND-95.3-LP-10, "ES-1.3, TRANSFER TO COLD LEG RECIRCULATION," rev 11 dtd 07/16/07.

6. Surry Procedure 1-ES-1.4, "TRANSFER TO HOT LEG RECIRCULATION," rev. 5.

7. Surry Lesson Plan ND-95.3-LP-11, "ES-1.4, TRANSFER TO HOT LEG RECIRCULATION," rev. 10 dtd 07/17/07.

References Provided to Applicant None.

Answer: C

5. 000000000022AA1.07 1 Initial conditions:

- Unit 1 is at 100% power.

- Letdown was removed from service to repair a leak.

- Excess letdown is in service.

- Pressurizer (PRZR) is at program level.

Current conditions:

AP-8.00, Loss of Normal Charging Flow, has been initiated.

- Charging flow and discharge pressure is erratic.

- Charging pump amps are erratic on the operating pump.

- Charging pump motors appear to be operating normally.

- System valve positions appear to be normal.

Based on the current conditions, which one of the following describes...

(1) how excess letdown flow will be affected and (2) whether a reactor trip is required in accordance with 1-AP-8.00?

A. (1) Excess letdown will continue until manually isolated.

(2) Reactor trip is NOT required.

B. (1) Excess letdown will be automatically isolated.

(2) Reactor trip is NOT required.

C. (1) Excess letdown will be automatically isolated.

(2) Reactor trip is required.

D. (1) Excess letdown will continue until manually isolated.

(2) Reactor trip is required.

K/A

Loss of Reactor Coolant Makeup Ability to operate and/or monitor Excess Letdown isolation containment valve switches and indicators as it applies to Loss of Reactor Coolant Makeup.

(CFR: 41.7/45.5/45.6) (RO - 2.8)

K/A Match Analysis The RO applicant is required to determine how excess letdown flow path is affected by a loss of Reactor makeup.

Answer Choice Analysis A. INCORRECT. Plausible because the first half of the response is correct - the excess letdown system will continue to operate. Also, if the applicant believes the conditions indicate something other than gas binding (e.g., low suction pressure, low flow, etc.) then other actions are taken to re-establish charging flow.

B. INCORRECT. Plausible because normal letdown will automatically isolate and the AP does direct a reactor trip if gas binding exists on the charging system.

C. INCORRECT. Plausible because normal letdown will automatically isolate. Also, if the applicant believes the conditions indicate something other than gas binding (e.g.,

low suction pressure, low flow, etc.) then other actions are taken to re-establish charging flow.

D. CORRECT. Excess letdown flow will continue until manually isolated and the AP does direct a reactor trip if gas binding exists on the charging system.

Supporting References

1. Surry lesson plan ND-88.3-LP-3, "Seal Injection," rev. 10, Obj E, p. 15.

2. 1-AP-8.00, Loss of Normal Charging, rev. 12, pg. 2

3. This question is modified from Sequoyah 2010-301 NRC exam (Q #5). Added "gas binding" conditions to the stem and modified the second half of the distractors to focus on whether a reactor trip is required versus identifying a condition that requires a reactor trip.

References Provided to Applicant None.

Answer: D

6. 000000000025AA1.12 1 Unit 1 Initial Sequence of events at 180 hours0.00208 days <br />0.05 hours <br />2.97619e-4 weeks <br />6.849e-5 months <br /> after shutdown:

- Mid loop operation was entered after a shutdown from 100%

- The running RHR pump suction was vortexing AP-27.00, LOSS OF DECAY HEAT REMOVAL CAPABILITY, was entered

- The crew reduced RHR flow and raised RCS level, which stopped the vortexing Current Conditions:

- RCS temperature is 145oF and increasing slowly

- The RHR controls are in their normal configuration Which one of the following correctly describes: (1) actions that will stabilize temperature, and (2) the response of RHR cold leg return flow after the actions have been taken?

A. (1) Manually open 1-HCV-1758 using the control switch.

(2) RHR cold leg return flow indication will rise.

B. (1) Manually open 1-HCV-1758 using the control switch.

(2) RHR cold leg return flow indication will remain approximately stable.

C. (1) Raise the setpoint on the controller to open 1-HCV-1758.

(2) RHR cold leg return flow indication will rise.

D. (1) Raise the setpoint on the controller to open 1-HCV-1758.

(2) RHR cold leg return flow indication will remain approximately stable.

K/A Loss of RHR Ability to operate and/or monitor the following as they apply to the loss of RHR System: RCS temperature indicators.

K/A Match Analysis Applicant must know how RHR flow will respond to the manipulations to control temperature.

Answer Choice Analysis A. INCORRECT. Incorrect. Plausible because, opening valves typically results in flow going up; however, in this system, there is another valve downstream of two parallel flow paths the maintains flow constant.

B. CORRECT. According to the LP, HCV-1758 has no automatic controls.

Therefore, this valve must be manually controlled. With controls in their normal configuration, FCV-1605 would be in auto controlling at the chosen setpt. Therefore,

no change in RHR flow should occur.

C. INCORRECT. 1-HCV-1758 does not have any auto functions; however, FCV-1605 does, which makes this choice plausible. Second part plausible for same reason as listed in A.

D. INCORRECT. See above.

Supporting References ND-88.2-LP-1, Residual Heat Removal System Description, Rev 8.

1-AP-27.00, LOSS OF DECAY HEAT REMOVAL CAPABILITY, Rev. 20.

References Provided to Applicant None.

Answer: B

7. 000000000026AA2.03 1 Unit 1 initial conditions:

- B RHR pump is running with RHR 1B heat exchanger aligned for Shutdown Cooling.

- Component Cooling (CC) pump,1-CC-P-1A, is running.

- The following annunciators are lit:

VSP D7, CC SURGE TK HI-LO LVL K E7, CC PPS DISCH HDR LO PRESS

- FI-1-100B, CC Supply Hdr Flow - Hdr B, has increased.

- FI-1-110B, RHR HX B CC Outlet Hdr B Flow, is stable.

- The NORM-CHILL switch is in the CHILL position.

- CC Surge Tank level is dropping.

- Crew has entered ARP 1-K E7.

Current conditions:

- In accordance with ARP 1-K E7 the following actions are complete:

- All CC pumps are in PTL.

- Charging and letdown have been secured.

- CC Surge Tank level is still dropping.

- Actions are in progress to locally establish makeup to the CC system and isolate the leak.

Based on the current conditions, which one of the following components requires isolation?

Close the CC inlet and outlet valves to the

A. Containment Air Recirculation fans.

B. Containment Instrument Air compressor seal coolers.

C. 'B RHR pump seal cooler.

D. RHR High Radiation Sample System Cooler.

K/A Loss of Component Cooling Water Ability to determine and/or interpret the valve lineups necessary to restart the CCWS while bypassing the portion of the system causing the abnormal condition as it applies to Loss of Component Cooling Water.

(CFR: 43.5/45.13) (RO - 2.6)

K/A Match Analysis The RO applicant must evaluate current plant configuration, CC flow indications and CC surge tank level trends to determine which component has experienced a CC leak and identify the component that should be isolated.

Answer Choice Analysis A. INCORRECT. Plausible because with the Containment Air Recirculation Fans (CARFs) aligned to the NORM position, it would result in a flow increase on FI-1-100B with no change in flow FI-1-110B (which is on return line from the RHR B heat exchanger) and a decrease in the CC surge tank level. However with the CARFs aligned to the CHILL position CC is isolated to the CARFs and cooling is provided by the Chilled CC system.

B. CORRECT. CC cooling is provided to the Containment Instrument Air compressor seal coolers from the B CC header. This would result in a flow increase on FI-1-100B with no change in flow FI-1-110B (which is on the return line from the RHR B heat exchanger) and a decrease in the CC surge tank level.

C. INCORRECT. Plausible because with RHR providing shutdown cooling a CC leak on the in-service RHR seal cooler would cause FI-1-100B to increase along with a drop in CC surge tank level. However, FI-110B would also indicate a flow increase if a CC leak existed on the RHR seal cooler.

D. INCORRECT. Plausible because the HRSS cooler for the RHR would be in service.

However, these coolers are supplied from the Auxiliary Bldg East header which taps off upstream of the B CC header and would not cause an indicated flow increase on

FI-1-100B.

Supporting References

1. Surry lesson plan ND-88-5-LP-1, "Component Cooling Systems," rev. 21, Objective B, slides 19, 25, 26, 30, 43 and 45.

2. 1K-E7, CC PPS DISCH HDR LO PRESS, Rev. 1

3. Dwg # 11448-FM-072A, Sh 1 of 7, Rev. 28

4. Dwg # 11448-FM-072B, Sh 1 of 3, Rev. 29

5. Dwg # 11448-FM-072B, Sh 2 of 3, Rev. 29*

6. Dwg # 11448-FM-072C, Sh 1 of 5, Rev. 39

7. Dwg # 11448-FM-072C, Sh 2 of 5, Rev. 34*

8. Dwg # 11448-FM-072D, Sh 2 of 5, Rev. 33

• Need to include copies of these drawings in the documentation file.

References Provided to Applicant None.

Answer: B

8. 000000000027AK2.03 1 Unit 1 Initial Conditions:

- Time = 1500.

- 100% Power.

- Pressurizer (PZR) Pressure = 2235 psig.

- All PZR Pressure control components are in their normal 100% power alignments.

Current Conditions:

- Time = 1502.

RC-PC-1444J, PZR Pressure Master Controller output fails to a constant value of 70%.

- No operator actions have been taken.

Based upon the current conditions, what is the expected configuration of the PZR Pressure control system?

A. Proportional Heaters OFF, Both Spray Valves FULL OPEN, PZR PORV PCV-1455C OPEN.

B. Proportional Heaters OFF, Both Spray Valves FULL OPEN, PZR PORV PCV-1455C CLOSED.

C. Proportional Heaters 50% OUTPUT, Both Spray Valves CLOSED, PZR PORV PCV-1455C CLOSED.

D. Proportional Heaters 100% OUTPUT, Both Spray Valves CLOSED, PZR PORV PCV-1455C CLOSED.

K/A Pressurizer Pressure Control System Malfunction 027AK2.03 Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the following: Controllers and positioners.

(CFR 41.7 / 45.7) (RO - 2.6)

K/A Match Analysis This question matches the K/A statement by requiring the applicant to correctly recall the relationship between master PZR pressure controller output and the expected response of the PZR pressure control system. The question is testing fundamental knowledge in that an applicant could arrive at the correct answer by straightforward recall of controller setpoints (such as the table on handout page 5.4 of the included Surry lesson plan). If the applicant does not recall this relationship directly, he or she could also arrive at the correct answer by calculation of pressure setpoints vs. percent controller output. To do so, the controller span (i.e. 0% to 100% is equivalent to 200 psig, -60 psig to +140 psig) must be remembered:

x 70%

=

200 100%

x = 140 x ~ NOP + 80 psig x = (2235 + 80) = 2315 psig .

Answer Choice Analysis A. INCORRECT. This distractor is plausible if an applicant mis-remembers the controller output setpoints from the PZR Pressure master controller. This distractor would correspond to a controller output of 80% or greater.

B. CORRECT. At a constant output of 70% a short time after PZR Pressure is at NOP (2235 psig), it is expected that the proportional (modulating) heaters would be OFF, the sprays OPEN, and the PORV CLOSED.

C. INCORRECT. This distractor is plausible if an applicant mis-remembers the controller output setpoints from the PZR Pressure master controller. This distractor

would correspond to a controller output of 30%--i.e, the applicant mentally failed the controller in the wrong Pressure direction.

D. CORRECT. This distractor is plausible if an applicant mis-remembers the controller output setpoints from the PZR Pressure master controller. This distractor would correspond to a controller output of 22.5% or lower--i.e. the applicant mentally failed the controller in the wrong Pressure direction.

Supporting References

1. Surry Lesson Plan ND-93.3-LP-5, Pressurizer Pressure Control, rev. 13 dtd 12/14/08. Specifically p. 7 and the summary of setpoints on handout 5.4.

References Provided to Applicant None.

Answer: B

9. 000000000029EA2.08 1 Unit 1 initial conditions:

- Reactor power = 12%

- Main Turbine surveillance testing is in progress

- IRPI: Bank A = 220 Steps Bank B = 220 Steps Bank C = 220 Steps Bank D = 140 Steps Current plant conditions:

- Main Turbine trip occurs

- IRPI:

Bank A = 220 Steps Bank B = 220 Steps Bank C = 220 Steps Bank D = 140 Steps except CR-XX which = 0 Steps Based on the above conditions, which one of the following states: (1) what the Control Rod Group Counter for Group D CR-XX should read and (2) the correct procedure the crew is required to initiate to address the event?

A. (1) 140 Steps (2) 1-E-0 REACTOR TRIP OR SAFETY INJECTION

B. (1) 140 Steps (2) 0-AP-1.00 CONTROL ROD SYSTEM MALFUNCTION C. (1) 0 Steps (2) 1-E-0 REACTOR TRIP OR SAFETY INJECTION D. (1) 0 Steps (2) 0-AP-1.00 CONTROL ROD SYSTEM MALFUNCTION K/A Ability to determine or interpret the following as they apply to an ATWS. Rod bank step counters and RPI.

K/A Match Analysis Requires applicant to know how equipment is effected by a fire based on the location of the fire.

Answer Choice Analysis A. Correct: Control Rod Group Counter will remain at its prior position. With reactor power = 12%, the reactor should have tripped (ATWS) so entry into E-0 is required.

B. Incorrect: 1st part is correct. The reactor should have tripped so E-0 is the correct procedure to enter. 2nd part is plausible because if power was < 10%, it would be correct.

C. Incorrect: Step counter does not track actual CR position. Plausible because it does show the demand signal for that CR. 2nd part is correct.

D. Incorrect: 1Step counter does not track actual CR position. Plausible because it does show the demand signal for that CR. The reactor should have tripped so E-0 is the correct procedure to enter. 2nd part is plausible because if power was < 10%, it would be correct.

Supporting References ND-93.3-LP3 Rod Control System Obj: K ND-93.3-LP4 CERPI System Obj: E References Provided to Applicant none Request Licensee to provide valid CR number and operation of step counters.

Answer: A

10. 000000000038EK1.02 1

Unit 1 Initial Conditions:

- 100% Power.

- Reactor Coolant System (RCS) Pressure is 2235 psig and stable.

- Steam Generator (S/G) 'A' Pressure is 790 psig and stable.

- A seismic event causes a complete shear of one tube in S/G 'A' coincident with a loss of offsite power.

- Initial tube rupture flow rate from the RCS to S/G 'A' is 800 gpm.

Current conditions:

- Operators correctly perform the step to isolate ruptured S/G(s) in 1-E-3, STEAM GENERATOR TUBE RUPTURE, and have properly adjusted the 'A' S/G PORV controller setpoint.

- S/G 'A' pressure is 20 psig lower than the 'A' S/G PORV controller setpoint.

- S/G 'A' PORV is closed.

- RCS pressure is 1650 psig.

Based on the current conditions, which one of the following is: (1) the current tube rupture flow rate from the RCS to S/G 'A,' AND (2) if S/G 'A' PORV were to open, the required action in accordance with 1-E-3 is to __________ ?

A. (1) 350 gpm (2) place 'A' S/G PORV controller in MANUAL and close the PORV.

B. (1) 350 gpm (2) verify 'A' S/G PORV closed when 'A' S/G pressure is less than setpoint.

C. (1) 530 gpm (2) place 'A' S/G PORV controller in MANUAL and close the PORV.

D. (1) 530 gpm (2) verify 'A' S/G PORV closed when 'A' S/G pressure is less than setpoint.

K/A Steam Generator Tube Rupture (SGTR) 038EK1.02 Knowledge of the operational implications of the following concepts as they apply to the SGTR: Leak rate vs. pressure drop.

(CFR 41.8 / 41.10 / 45.3) (RO - 3.2)

K/A Match Analysis This question matches the K/A statement by requiring the applicants to correctly calculate a current SGTR leak rate, given initial and current conditions that correspond to the design basis SGTR accident as described in the Surry FSAR. To arrive at the

correct answer, the applicant must recall/understand the basic relationship between volumetric flow rates and differential pressure. The applicant must also recall an important S/G PORV pressure setpoint from E-3 in order to obtain the correct flow rate value. The second part of the question requires the applicant to correctly recall the required actions for an open ruptured S/G PORV, in accordance with 1-E-3.

Answer Choice Analysis A. INCORRECT. This distractor is based upon an incorrect belief that volumetric flowrate is directly proportional to pressure drop (instead of the square root of differential pressure) multiplied by a density correction factor. This mistake is plausible because it is approximately the method used to determine mass flow rates. 1-E-3, STEAM GENERATOR TUBE RUPTURE, requires that the ruptured S/G PORV controller setpoint be adjusted to 1035 psig as part of the step to isolate the ruptured S/G (step 3). Because the question stem states that the ruptured S/G pressure is 20 psig less than the setpoint, the applicant will use 1015 psig in the calculation, which proceeds as follows:

V RCS P (incorrect formula )

47.03482 lbm 3 (2235 psig 790 psig )

800 gpm ft

=

46.63061 lbm 3 (1650 psig 1015 psig )

Vcurrent ft 800 gpm 67965.315

=

Vcurrent 29610.43735 Vcurrent = 349 gpm (incorrect )

The second part of this distractor is also incorrect, because step 3.b. RNO states WHEN ruptured SG pressure less than 1035 psig, THEN verify SG PORV closed.

The second part distractor is plausible, because it is the required action if the SG PORV does not close when pressure goes less than 1035 psig. It is further plausible because closing the PORV will terminate a radioactive release, and also the closed PORV will create a higher backpressure, which will lead to a lower SGTR flow rate from the RCS to the ruptured SG (K/A match).

B. INCORRECT. This distractor is based upon an incorrect belief that volumetric flowrate is directly proportional to pressure drop (instead of the square root of differential pressure) multiplied by a density correction factor. This mistake is plausible because it is approximately the method used to determine mass flow rates. 1-E-3, STEAM GENERATOR TUBE RUPTURE, requires that the ruptured S/G PORV controller setpoint be adjusted to 1035 psig as part of the step to isolate the ruptured S/G (step 3). Because the question stem states that the ruptured S/G pressure is 20

psig less than the setpoint, the applicant will use 1015 psig in the calculation, which proceeds as follows:

V RCS P (incorrect formula )

47.03482 lbm 3 (2235 psig 790 psig )

800 gpm ft

=

46.63061 lbm 3 (1650 psig 1015 psig )

Vcurrent ft 800 gpm 67965.315

=

Vcurrent 29610.43735 Vcurrent = 349 gpm (incorrect )

The second part of this distractor is correct, because step 3.b. RNO states WHEN ruptured SG pressure less than 1035 psig, THEN verify SG PORV closed.

C. INCORRECT. 1-E-3, STEAM GENERATOR TUBE RUPTURE, requires that the ruptured S/G PORV controller setpoint be adjusted to 1035 psig as part of the step to isolate the ruptured S/G (step 3). Because the question stem states that the ruptured S/G pressure is 20 psig less than the setpoint, the applicant will use 1015 psig in the calculation, which proceeds as follows:

V P 800 gpm (2235 psig 790 psig )

=

Vcurrent (1650 psig 1015 psig )

800 gpm 38.0131556

=

Vcurrent 25.1992063 Vcurrent = 530 gpm (correct answer ) q.e.d .

The second part of this distractor is incorrect, because step 3.b. RNO states WHEN ruptured SG pressure less than 1035 psig, THEN verify SG PORV closed. The second part distractor is plausible, because it is the required action if the SG PORV does not close when pressure goes less than 1035 psig. It is further plausible because closing the PORV will terminate a radioactive release, and also the closed PORV will create a higher backpressure, which will lead to a lower SGTR flow rate from the RCS to the ruptured SG (K/A match).

D. CORRECT. 1-E-3, STEAM GENERATOR TUBE RUPTURE, requires that the

ruptured S/G PORV controller setpoint be adjusted to 1035 psig as part of the step to isolate the ruptured S/G (step 3). Because the question stem states that the ruptured S/G pressure is 20 psig less than the setpoint, the applicant will use 1015 psig in the calculation, which proceeds as follows:

V P 800 gpm (2235 psig 790 psig )

=

Vcurrent (1650 psig 1015 psig )

800 gpm 38.0131556

=

Vcurrent 25.1992063 Vcurrent = 530 gpm (correct answer ) q.e.d .

The second part of this distractor is also correct, because step 3.b. RNO states WHEN ruptured SG pressure less than 1035 psig, THEN verify SG PORV closed.

Supporting References

1. Surry Power Station UFSAR rev. 41 (as of 09/20/09) section 14.3.1, Steam Generator Tube Rupture. On p. 14.3-4, under figure 14.3-6, Break Flow, it is stated that initial break flow is approximately 800 gpm, the value used in the question stem.

2. Surry Procedure 1-E-3, STEAM GENERATOR TUBE RUPTURE, rev. 38. Step 3 of 1-E-3, ISOLATE RUPTURED SG(s): requires the ruptured S/G PORV controller setpoint be adjusted to 1035 psig.

3. Surry Lesson Plan ND-89.1-LP-2 rev. 24, Main Steam System. Page 19 states that main steam pressure is approximately 800 psig at 100% power (close to the 790 psig used in the question stem).

4. Steam Tables. http://www.steamtablesonline.com. Printout provided (density values for initial and current conditions).

References Provided to Applicant Steam Tables.

Answer: D

11. 000000000054AK1.02 1

Unit 1 Initial Conditions:

• 1-FR-H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK, is in progress

• RCS Bleed and Feed has been established Unit 1 Current Conditions:

• Hot leg RCS temperatures are 500oF and stable

• CETC temperatures are 500oF and stable (Have licensee adjust the above temps down to something realistic for Bleed and Feed)

• SG WR Levels are: A = 4%, B = 6%, C = 5%

• Containment pressure is 25 psia and lowering Based on the current conditions, which one of the following states the 1-FR-H.1 requirements for establishing AFW?

A. Feed SGs at a minimum total flow of 350 gpm.

B. Feed SGs at a minimum total flow of 450 gpm.

C. Feed only one SG at no more than 60 gpm.

D. Feed only one SG at no more than 100 gpm.

K/A Loss of Main Feedwater Knowledge of the operational implications of the following concepts as they apply to Loss of Main Feedwater (MFW): Effects of feedwater introduction on dry S/G.

K/A Match Analysis The question requires the applicant to know the operational implications (I.E. AFW feed requirements) of feeding a dry steam generator.

Answer Choice Analysis A. INCORRECT. Plausible because the only difference between this and the correct answer is that this choice uses the non-adverse value for AFW flow requirements.

B. CORRECT. Adverse containment numbers are in effect because ctmt pressure is above 20 psia. Step 23 of H.1 directs that attempts to establish AFW flow should still be made iaw Step 2. Step 2 states that total flow should be at least 450 gpm.

C. INCORRECT. See D below. The same justification exists for C, except that the non-adverse value is used.

D. INCORRECT. The Continuous Action Page criteria are only to be used for a hot/dry SG, which is indicated by WR SG levels < 7%[22%] AND RCS hot leg temperatures greater than 550 F. The stem of this question has temperatures at less than 550 F. If the applicant applied the criteria for a hot dry SG, then this would be a correct answer.

Supporting References

1. 1-FR-H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK, Rev. 28.

References Provided to Applicant None.

Answer: B

12. 000000000057AG2.2.37 1 Unit 1 Initial Conditions:

• Power = 100%.

• Time = 0255. A large electrical fire at the site caused the loss of Vital Buses 1-III, 1-IIIA, 1-IV and 1-IVA.

Current conditions:

• Time = 0300. Operators initiate a manual Reactor Trip.

• Time = 0301. One Reactor Coolant Pump (RCP) is secured.

• Time = 0303. Operators note the following:

- Loop A Tave = 541 °F

- Loop B Tave = 551 °F

- Loop C Tave = 552 °F

• Time = 0304. Safety Injection is NOT actuated.

Based upon the current conditions, which one of the following describes (1) whether an automatic safety injection should have occurred by Time = 0304, AND (2) how many RCPs have lost Closed Cooling (CC) to the lube oil coolers?

A. (1) An automatic Safety Injection should have actuated.

(2) One RCP has lost CC to the lube oil coolers.

B. (1) An automatic Safety Injection should NOT have actuated.

(2) More than one RCP has lost CC to the lube oil coolers.

C. (1) An automatic Safety Injection should have actuated.

(2) More than one RCP has lost CC to the lube oil coolers.

D. (1) An automatic Safety Injection should NOT have actuated.

(2) One RCP has lost CC to the lube oil coolers.

K/A 057 Loss of Vital AC Instrument Bus 057AG2.2.37 Ability to determine operability and/or availability of safety related equipment.

(CFR 41.7 / 43.5 / 45.12) (RO - 3.6)

K/A Match Analysis This question matches the K/A statement by requiring the RO applicant to determine whether a safety injection should have occurred, and the status of vital CC cooling to RCP motors, given an operationally valid scenario where multiple Vital AC Instrument Buses have lost power.

Answer Choice Analysis A. INCORRECT. Part (1) of this distractor is correct. The Surry lesson plan for Vital and Semi-Vital Bus Distribution states the following for a loss of Vital Bus III: It should be noted that Safety Injection is imminent if A or B Loop Tave drops below 543 °F.

This is due to the High Steam Flow/Low Tave Channel III bistables being in a tripped condition due to loss of power. A NOTE in 1-AP-10.03, LOSS OF VITAL BUS III, before step 7 states essentially the same thing as the lesson plan. Therefore, based on the conditions given in the stem, SI should have automatically actuated. Part (2) of this distractor is incorrect. The Surry lesson plan again states: A loss of Vital Bus II, III, or IV requires the operator to manually trip the reactor and secure the appropriate RCP due to a loss of CC to the lube oil coolers for the RCP. Because both Vital Bus III and IV have been lost, CC to the lube oil coolers has been lost to the A and C RCPs.

B. INCORRECT. Part (1) is incorrect, part (2) correct. See analysis of A above.

C. CORRECT. Parts (1) and (2) correct. See analysis of A above.

D. INCORRECT. Part (1) incorrect, part (2) incorrect. Plausible if an applicant wrongly remembers the power supply/Vital Bus arrangement to the High Steam Flow/Low Tave SI bistables; also plausible because if Vital Bus I is lost, CC to the lube oil coolers to a RCP is not lost (if an applicant mis-remembers the correct Vital Bus lineup).

Supporting References

1. Surry Procedure 1-AP-10.03, LOSS OF VITAL BUS III, rev. 16. Especially NOTE before step 7.

2. Surry Lesson Plan ND-90.3-LP-5, VITAL AND SEMI-VITAL BUS DISTRIBUTION, rev. 17.

References Provided to Applicant None.

Answer: C

13. 000000000058G2.2.36 1 Initial conditions:

• Unit 1 is at 100% power.

• The semi-annual Station Battery Discharge Test for the 1B 125 VDC bus is in progress.

• Uninterruptible Power Supplies (UPS) 1B1 and 1B2 were de-energized at 1035 hrs and discharge of the 1B battery bank commenced.

Current conditions:

• Output breaker for 1A battery bank tripped during performance of the test.

Which one of the following describes the status of LCO 3.16, Emergency Power System, during testing AND LCO 3.9, Station Service Systems, after trip of the output breaker for 1A battery bank?

During Test - Prior to Breaker Trip After Breaker Trip A. Actions of LCO 3.16 are NOT required. Actions of LCO 3.9 are NOT required.

B. Actions of LCO 3.16 are NOT required. Actions of LCO 3.9 are required.

C. Actions of LCO 3.16 are required. Actions of LCO 3.9 are NOT required.

D. Actions of LCO 3.16 are required. Actions of LCO 3.9 are required.

K/A Loss of DC Power Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations. (as related to Loss of DC Power)

(CFR: 41.10/43.2/45.13) (RO - 3.1)

K/A Match Analysis The RO applicant must determine how an off-normal alignment of the 1B 125 VDC system to support surveillance testing (preventive maintenance) effects the LCOs status during the test and after the 1A Battery Bank output breaker trips.

Answer Choice Analysis NOTE: The LCOs of interest are:

(1) LCO 3.16.A.6, which requires two station batteries, two battery chargers (UPSs) and the DC Distribution systems to be OPERABLE, and (2) LCO 3.9.A.3 and A.4 which require both DC buses to be energized and one battery charger (UPS) per battery operating, respectively as explained in LCO 3.16.

The surveillance test de-energizes both UPSs. However, Technical Specification Interpretations, specifically TSI-015C, Semi-annual Station Battery Discharge Test states that when this surveillance is being performed the station battery chargers can be considered operable and entry into an action statement is not required. It specifically addresses 3.16.A.6 and 3.9.A.4 A. CORRECT. Based on TSI-015C, actions for LCO 3.16 are not required during the test. Once the 1A battery bank output breaker trips, both DC buses remain energized (1A bus from the two UPSs and 1B bus from the 1B Battery Bank) and per TSI-015C there are still two UPSs operable on the 1A bus and two UPSs operable on the 1B bus.

Therefore, no actions are required for LCO 3.9.

B. INCORRECT. Plausible because based on TSI-015C, the first half of the response is correct. For the second half of the response, if the applicant fails to apply TSI-015C then it would appear that the conditions for LCO 3.9. are not met (i.e., one operating UPS per battery operating and the DC Distribution systems to be energized).

However, both DC Distribution systems are still energized (1A bus by the battery chargers and 1B bus by the battery bank and the UPSs for 1B bus can be re-energized).

C. INCORRECT. Plausible because if the applicant fails to apply TSI-015C, LCO 3.16. requires two station batteries, two UPSs and the DC Distribution systems to be

OPERABLE. When the test de-energizes the UPSs the applicant may assume that an action statement for LCO 3.16 should be entered. The second half of the response is correct (See discussion in choice A above.).

D. INCORRECT. Plausible because if the applicant fails to apply TSI-015C, LCO 3.16.

requires two station batteries, two UPSs and the DC Distribution systems to be OPERABLE. When the test de-energizes the UPSs the applicant may assume that an action statement for LCO 3.16 should be entered. For the second half of the response, if the applicant fails to apply TSI-015C then it would appear that the conditions for LCO 3.9. are not met (i.e., one operating UPS per battery operating and the DC Distribution systems to be energized). However, both DC Distribution systems are still energized (1A bus by the battery chargers and 1B bus by the battery bank and the UPSs for 1B bus can be re-energized).

Supporting References

1. Surry lesson plan ND-90.3-LP-6, "125VDC DISTRIBUTION," rev. 14, Obj. E, p. 4.

2. TSI-015C, Semi-annual Station Battery Discharge Test, 10/19/2000.

3. Technical Specifications, LCO 3.9 and LCO 3.16, pg 3.9-1 and 3.16-2 References Provided to Applicant None.

Answer: A

14. 000000000062AK3.04 1 The following Unit 1 conditions exist:

Power is 100%

Maintenance activities cause 1-SW-MOV-102A and B, Component Cooling Supply Valves to throttle almost entirely closed.

Which one of the following describes how closing 1-SW-MOV-102A and B will impact the SW temperatures at the discharge of the BC Water Coolers and the reason?

A. Temperatures will be unaffected because the BC Water Coolers are in parallel with 1-SW-MOV-102A and B and temperatures are maintained by TCVs.

B. Temperatures will be unaffected because of the small magnitude of the heat load from the BC Water Coolers.

C. Temperatures will change because of increased flow through the BC Water Coolers.

D. Temperatures will change because of decreased flow through the BC Water Coolers.

K/A Loss of Nuclear Service Water Knowledge of the reasons for the following responses as they apply to the Loss of Nuclear Service Water: Effect on the nuclear service water discharge flow header of a loss of CCW.

K/A Match Analysis The loss of CCW occurs with the closure of 102A and B, which provide cooling to the CCW heat exchangers. The applicant then is required to know how this affects the SW discharge header. Flow through the BC Coolers is part of the discharge header. The applicant must know the reason for why temperatures at the exit of the coolers are changing.

Answer Choice Analysis A. INCORRECT. I did not see any information in the training material that would suggest that TCVs would act to maintain temperature. Plausible due to TCVs being used in conjunction with other heat exchangers in the plant.

B. INCORRECT. The increased flow through the heat exchanger would produce a visible change in Cooler outlet temperature. Plausible because the applicant may believe the heat loads to be small.

C. CORRECT. Closing off flow to the CCW HX will force more flow through the BC Coolers, thereby lowering the discharge temps due to the increased flow.

D. INCORRECT. See above. Plausible if the applicant thinks the BC Coolers are downstream of MOV-102A and B.

Supporting References LP-ND-89.5-2, Service Water System, Rev. 16.

References Provided to Applicant None.

Answer: C

15. 000000000077AK3.02 1

Current conditions:

• The station has been notified that the real-time contingency analysis program is unavailable.

• Unit 1 and Unit 2 crews have entered 0-AP-10.18, Response to Grid Instability.

• While performing step #7 (Check system conditions), the system operator reports the following voltage readings:

• 230 KV bus = 242 KV

• 500 KV bus = 503 KV Which one of the following describes:

(1) the concern with these readings in accordance with 0-AP-10.18 and (2) the reason for the concern?

A. (1) The maximum voltage limit has been exceeded on the 230 KV bus.

(2) Over an extended period of time the high voltage condition damages energized equipment on Unit 1.

B. (1) The maximum voltage limit has been exceeded on the 230 KV bus.

(2) In the event of a Unit 1 trip and Engineered Safety Feature actuation, the high voltage condition damages station bus breakers to the point where only the emergency diesel generators are available to supply the emergency buses.

C. (1) The minimum voltage limit has been reached on the 500 KV bus.

(2) Over an extended period of time the low voltage condition damages energized equipment on Unit 2.

D. (1) The minimum voltage limit has been reached on the 500 KV bus.

(2) In the event of a Unit 1 trip and Engineered Safety Feature actuation, the low voltage condition degrades station bus voltage to the point where the emergency diesel generators are needed to supply the emergency buses.

K/A Generator Voltage and Electric Grid Disturbances Knowledge of the reasons for the following responses as they apply to Generator Voltage and Electric Grid Disturbances: Actions contained in abnormal procedure for voltage and grid disturbances. (as related to Loss of DC Power)

(CFR: 41.4, 41.5, 41.7, 41.10/45.8) (RO - 3.6)

K/A Match Analysis

The RO applicant is required to recognize off-normal voltage conditions on the offsite sources addressed in the abnormal procedure for grid disturbances and understand the reason for the voltage limits contained in the abnormal procedure.

Answer Choice Analysis A. INCORRECT. Plausible because the voltage is significantly higher than the schedule deviation (+/-2 KV) allowed by 1-OP-26.5, 230 KV Switchyard Voltage.

However, the actual maximum voltage in accordance with 0-AP-10.18 is 245 KV on the 230 KV bus. The reason for the concern is correct for exceeding the maximum voltage.

B. INCORRECT. Plausible because the voltage is significantly higher than the schedule deviation (+/-2 KV) allowed by 1-OP-26.5, 230 KV Switchyard Voltage.

However, the actual maximum voltage in accordance with 0-AP-10.18 is 245 KV on the 230 KV bus. The reason for the concern is plausible because the emergency buses being supplied from the EDGs is a real concern but it is not due to a high voltage condition but as a result of degraded voltage to the station buses.

C. INCORRECT. Plausible because the minimum voltage limit has been reached on the 500 KV bus. The reason for the concern is plausible because Unit 2 is normally aligned to the 500KV bus and this is a valid concern for a high voltage condition, but it is not the reason for a low voltage condition.

D. CORRECT. The minimum voltage limit on the 500 KV bus is 505 KV. The limit is due to a concern that below the minimum voltage of 505 KV, the normal supply breakers will open and will result in the EDGs carrying all the loads on the emergency buses following a unit trip and start of all the safety-related equipment.

Supporting References

1. Surry lesson plan ND-95.1-LP-8, "Loss of Offsite Power," rev. 12, Objective C, pg.

18.

2. 0-AP-10.18, Response to Grid Instability, Rev. 19, pgs. 5 & 7 References Provided to Applicant None.

Answer: D

16. 000000000WE04EA1.3 1 Unit 1 Initial Conditions:

• An automatic Reactor Trip and Safety Injection occurred from 100% power.

Current Conditions:

• Operators have just transitioned to 1-ECA-1.2, LOCA OUTSIDE CONTAINMENT.

Which ONE of the following parameters is used to determine if the break has been isolated, in accordance with 1-ECA-1.2?

A. RVLIS level B. RCS subcooling C. RCS pressure D. PRZR level K/A LOCA Outside Containment W/E04EA1.3 Ability to operate and/or monitor the following as they apply to the (LOCA Outside Containment): Desired operating results during abnormal and emergency situations.

(CFR 41.7 / 45.5 / 45.6) (RO - 3.8)

K/A Match Analysis This question matches the K/A statement by requiring the applicant to correctly determine the proper system parameter used in ECA-1.2 to determine if the LOCA has been isolated by the actions of ECA-1.2. The question is a straightforward memory-level examination of exactly what the K/A requires us to test: desired operating results. Specifically, in ECA-1.2 the ultimate desired operating result (or outcome) is RCS pressure increasing due to isolating the break.

Answer Choice Analysis A. INCORRECT. This distractor is plausible because RVLIS level should begin to INCREASE once the LOCA is isolated and SI flow continues. However, RVLIS level is not the parameter required to be monitored by ECA-1.2, and A is therefore an incorrect distractor. Previous versions of this question used Pressurizer level for this distractor; however, the exam author changed this to RVLIS level due to his belief that it was a more plausible choice.

B. INCORRECT. This distractor is plausible because once the LOCA is isolated, SI flow and an increase in RCS pressure would also cause Subcooling to increase; however, this is not the parameter required by ECA-1.2 and is therefore incorrect.

C. CORRECT. The high-level action statement of step 2 of 1-ECA-1.2 is to TRY TO IDENTIFY AND ISOLATE BREAK, and step b) states Check RCS pressure -

INCREASING. If the answer is yes, the operator is eventually directed to transition to E-1 based on the success of the isolation efforts.

D. INCORRECT. This distractor is plausible because once the LOCA is isolated, Pressurizer level would be expected to increase due to SI injection flow.

Supporting References

1. Surry Procedure 1-ECA-1.2, LOCA OUTSIDE CONTAINMENT, rev. 6.

2. Surry Lesson Plan ND-95.3-LP-21, ECA-1.2, LOCA OUTSIDE CONTAINMENT, rev. 7 dtd 11/13/96.

3. This question is modified from Turkey Point W/E04EA2.2 from 2002-301 and Farley WE04EA2.2 from 2006-301 to apply to the Surry Power Station and to enhance plausibility.

References Provided to Applicant None.

Answer: C

17. 000000000WE05EK2.2 1 Unit 1 current conditions:

- 1-FR-H.1, Response to Loss of Secondary Heat Sink, has been initiated.

- All Main Feedwater (MFW) and Auxiliary Feedwater (AFW) pumps are unavailable.

- Unit 2 AFW cross-tie valve will not open.

- Charging and letdown are in service.

- Pressurizer pressure is 2330 psig.

- SG Wide range level indications are as follows:

- SG 1A = 10% SG 1B = 9% SG 1C =

10%

- The crew has reached step 5 of 1-FR-H.1 (Depressurize RCS to1950 psig).

Based on the current conditions, which one of the following is consistent with the mitigation strategy :

A. Initiate RCS Feed and Bleed immediately due to inadequate steam generator inventory.

B. Initiate RCS Feed and Bleed immediately due to excessive RCS pressure.

C. Continue with RCS depressurization to 1950 psig using one pressurizer PORV.

D. Continue with RCS depressurization to 1950 psig using auxiliary spray.

K/A Loss of Secondary Heat Sink Knowledge of the interrelations between the (Loss of Secondary Heat Sink) and the following: Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and the relations between proper operation of these systems to operation of the facility.

(CFR: 41.7/45.7) (RO - 3.9)

K/A Match Analysis The RO applicant is required to distinguish whether conditions require initiating RCS feed and bleed or continuing performance of the normal sequence of actions contained in FR-H.1, Loss of Secondary Heat Sink.

Answer Choice Analysis A. INCORRECT. Plausible because all the SGs are less than 12% level which is used as a level limit for other steps in the procedure. However, the 12% limit is based on the narrow range level gauges not the wide range gauges, which establish a level limit of 7% wide range as the point where RCS Feed and Bleed are to be initiated (See caution statement before step 2 and Foldout page of FR-H.1).

B. INCORRECT. Plausible because RCS pressure is abnormally high and FR-H.1 directs RCS Feed and Bleed due to high RCS pressure. However, the pressure limit is 2335 psig(See caution statement before step 2 and Foldout page of FR-H.1).

C. INCORRECT. Plausible because continuing with RCS depressurization to 1950 is the correct action. However, the procedure step directs the use of the pressurizer PORV only if letdown is not in service.

D. CORRECT. Continuing RCS depressurization to 1950 psig is the correct action

using the auxiliary spray to minimize thermal shock on the pressurizer.

Supporting References

1. Surry lesson plan ND-95.3-LP-41, "FR-H.1," rev. 12, Objective C, pgs. 17-18, 21.

2. 1-FR-H.1, Response to Loss of Secondary Heat Sink, rev. 8, step 5 and foldout page References Provided to Applicant none Answer: D

18. 000000000WE12EA2.2 1 Unit 1 initial plant conditions:

- Reactor power = 100%

- Main Steam Line break inside containment occurs

- Main Steam Line Isolation Valves failed to close

- Transition to 1-ECA-2.1, UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS has occurred Current plant conditions:

- SG A, B and C NR Level = 14%

- 20.7 psia Based on the above conditions, which ONE of the following states: (1) The Minimum AFW flow requirements for the SGs and (2) The maximum cooldown rate allowed IAW 1-ECA-2.1?

A. (1) 60 gpm (2) 100 oF/hr B. (1) 60 gpm (2) 50 oF/hr C. (1) 100 gpm (2) 100 oF/hr D. (1) 100 gpm (2) 50 oF/hr K/A Steam Line Rupture - Excessive Heat Transfer. Ability to determine and interpret the

following as they apply to the (Uncontrolled Depressurization of all Steam Generators):

Adherence to appropriate procedures and operation within limitations in the facilitys license and amendments.

K/A Match Analysis Question requires knowledge of facility procedures including cooldown rate limitations during a faulted SG.

Answer Choice Analysis A. INCORRECT. Per 10ECA-2.1, if SG NR level < 12% [18%], maintain a minimum of 60 gpm [100gpm]. Plausible because the minimum flow is 60 gpm if < 12% [no adverse containment conditions]. Cooldown rate is limited to 100 0F/hr.

B. INCORRECT. 1st part is plausible because the minimum flow is 60 gpm if <

nd 12% [no adverse containment conditions]. 2 part is plausible because for natural circ cooldown, the CD rate is limited to 50 0F/hr.

C. CORRECT. Per 10ECA-2.1, if SG NR level < 12% [18%], maintain a minimum of 60 gpm [100gpm]. With all 3 SGs blowing down in containment, adverse conditions will exist. Cooldown rate is limited to 100 0F/hr.

D. INCORRECT. 1st part is correct. 2nd part is plausible because for natural circ cooldown, the CD rate is limited to 50 0F/hr.

Supporting References 1-ECA-2.1, UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS ND-95.3-LP-22 Obj: A References Provided to Applicant None.

Answer:

19. 000000003AA1.02 1 Unit 1 Initial Conditions:

• 70% Power.

• 0-AP-1.01, CONTROL ROD MISALIGNMENT, is being implemented.

• A dropped Control Bank C rod has just been re-aligned with its group, but the lift coil disconnect switches have NOT been returned to normal.

Current Conditions:

• While attempting to operate the INTERNAL ALARM RESET pushbutton, the operator inadvertently operates the STARTUP RESET pushbutton.

Based on the current conditions, which one of the following describes the effect of operating the incorrect reset?

A. All of the Control Bank 'C' rods drop into the core. No other rods drop into the core.

B. All rods, including all Control Bank and Shutdown Bank rods, drop into the core.

C. All rods remain in their current position. There is no effect on the Rod Control System circuitry.

D. All rods remain in their current position. The Rod Control System circuitry indicates all rods are fully inserted.

K/A Dropped Control Rod 003AA1.02 Ability to operate and/or monitor the following as they apply to the Dropped Control Rod: Controls and components necessary to recover rod.

(CFR 41.7 / 45.5 / 45.6) (RO - 3.6)

K/A Match Analysis This question matches the K/A statement by testing the applicants ability to monitor the expected plant response during a dropped rod recovery, when an important component is mis-operated. The applicant needs to apply understanding of the rod control system, specifically the function of the rod control startup pushbutton, to arrive at the correct answer.

Answer Choice Analysis A. INCORRECT. This distractor is plausible if the applicant misunderstands the rod control system alignment during rod recovery, and believes that the lift coil disconnect switch alignment will cause only the Control Bank C rods to drop into the core.

Plausibility for this distractor is enhanced by emphasizing in the stem that the lift coil disconnect switches are still in the abnormal alignment required to recover the dropped rod.

B. INCORRECT. This distractor is plausible if the applicant does not understand the operation of the rod control startup pushbutton, and confuses the rod control startup

pushbutton with the Rx trip breaker reset pushbutton, and believes operation of this button cycles the reactor trip breakers (which would cause all rods to drop).

C. INCORRECT. The first part of this distractor is correct; no rods will drop when the rod control startup pushbutton is pressed. However, all rod indications will reset to zero, and therefore there will be an effect on Rod Control System circuitry.

D. CORRECT. As per Surry lesson plan ND-93.3-LP-3, Rod Control System, the STARTUP RESET pushbutton resets the following to zero: P/A converter counters, Internal memory and alarm circuit, group step counters, slave cycler counters, master cycler counters, and bank overlap counters. Therefore, no rods would move, and all indications would reset to zero.

Supporting References

1. Surry Procedure 0-AP-1.01, CONTROL ROD MISALIGNMENT, rev. 20.

especially step 27.

2. Surry Lesson Plan ND-95.3-LP-3 Rod Control System, rev. 19 dtd 09/01/09.

3. This question is modified from Harris 003AA1.02 to apply to the Surry Power Station and to enhance plausibility.

References Provided to Applicant None.

Answer: D

20. 000000005AG2.4.46 1 Current conditions:

• Reactor power = 46%

• The plant is stable following a turbine runback.

• Control bank D position indications are as follows:

Step Counter indications IRPI indications Group 1 rods - 188 steps Group 1 rods - 188 steps Group 2 rods - 166 steps Group 2 rods - 166 steps Based on the current conditions, which one of the following describes the status of the following alarms?

• 1G-A6, ROD CONT SYS URGENT FAILURE.

• 1G-B5, COMPUTER PRINTOUT ROD CONT SYS.

A. 1G-A6 NOT lit.

1G-B5 NOT lit.

B. 1G-A6 NOT lit.

1G-B5 lit.

C. 1G-A6 lit.

1G-B5 NOT lit.

D. 1G-A6 lit.

1G-B5 lit.

K/A Inoperable / Stuck Control Rod Ability to verify that the alarms are consistent with the plant conditions.

(CFR: 41.10/43.5/45.3/45.12) (RO - 4.2)

K/A Match Analysis The RO applicant is required to evaluate rod position information and determine that conditions would result in a rod urgent failure alarm.

Answer Choice Analysis A. INCORRECT. Plausible because no deviation appears between the associated groups step counter and the IRPI indications for the group.

B. INCORRECT. A COMPUTER PRINTOUT ROD CONT SYS alarm is plausible because at this power level a deviation of 20 steps or more between IRPI indication and the associated step counter will cause an actuation of this alarm. However, the deviation in position indication is between the two groups of rods not a deviation within the same rod group.

C. CORRECT. The deviation in position indication between the two rod groups would be consistent with an urgent rod failure condition (e.g. malfunction of the slave cycler for Group 1 rods).

D. INCORRECT. The first half of the answer is correct. In addition, a COMPUTER PRINTOUT ROD CONT SYS alarm is plausible because at this power level a deviation of 20 steps or more between IRPI indication and the associated step counter will cause an actuation of this alarm. However, the deviation in position indication is between the

two groups of rods not a deviation within the same rod group.

Supporting References

1. Surry lesson plan ND-93.3-LP-3, "Rod Control System," rev. 19, Obj F, pg. 60 - 61.

2. Annunciator Response Procedure, 1G-B5, COMPUTER PRINTOUT ROD CONT SYS, Rev. 14.

3. Annunciator Response Procedure, 1G-A6, ROD CONT SYS URGENT FAILURE, Rev. 1.

4. This question is modified from Harris 2008-301 Exam (Q #5). Modified conditions to be consistent with an urgent rod system failure rather than a dropped/mispositioned rod.

References Provided to Applicant None.

Answer: C

21. 000000032AK3.02 1 Initial Unit 1 conditions:

Reactor Power = 100%

Unit 1 current conditions:

A loss of all AC power has occurred Actions of 1-ECA-0.0, LOSS OF ALL AC POWER, are in progress SG ARVs are being controlled locally to reduce SG pressure to < 300 psig Both source-range nuclear instruments have failed low Containment pressure peaked at 21 psia and is now 19 psia Containment Radiation is 1.0E5 R/hr and slowly rising.

Based on the current conditions, which one of the following states:

(1) The nuclear instrument required to be monitored to check the reactor subcritical

and, (2) The reason for monitoring the nuclear instrument selected in part (1)?

A. (1) Intermediate range nuclear instruments.

(2) Adverse containment conditions exist due only to containment radiation.

B. (1) Intermediate range nuclear instruments.

(2) Adverse containment conditions exist due to both containment radiation and containment pressure.

C. (1) Wide range nuclear instruments.

(2) Adverse containment conditions exist due only to containment radiation.

D. (1) Wide range nuclear instruments.

(2) Adverse containment conditions exist due to both containment radiation and containment pressure.

K/A Loss of Source Range Nuclear Instrumentation Knowledge of the reasons for the following responses as they apply to the Loss of Source Range Nuclear Instrumentation: Guidance contained in EOP for loss of source-range nuclear instrumentation.

K/A Match Analysis The applicant is required to know which instruments to monitor when SR instruments have been lost and the reason why that instrument is required to be used.

Answer Choice Analysis A. (1) Incorrect - WR required IAW ECA-0.0 page 19 of 24.

(2) Incorrect. Adv CTMT numbers are required for both conditions. Even though pressure is now below the setpoint, once entered, the adv numbers cannot be exited due to the pressure drop. (VERIFY WITH LICENSEE)

B. (1) Incorrect. See above.

(2) Correct. See E-1 foldout page.

C. (1) Correct. See above.

(2) Incorrect. See above.

D. (1) Correct. See above.

(2) Correct. See above.

Supporting References 1-E-1, Loss of Secondary or Reactor Coolant, Rev. 34.

1-ECA-0.0, Loss of All AC Power, Rev. 32.

References Provided to Applicant None.

DISCUSS WITH THE LICENSEE IF ADV CTMT CAN BE EXITED IF EITHER PRESSURE OR RADIATION DROPS BACK BELOW THE SETPOINT FOR ADV NUMBERS.

Answer: D

22. 000000051AG2.4.35 1 Unit 1 Initial Conditions:

• 100% Power.

• The operations crew notes degrading condenser vacuum and enters 1-AP-14.00, "LOSS OF MAIN CONDENSER VACUUM."

• The crew is performing steps in Attachment 3, Low Air Ejector Flow Rate.

Current Conditions:

• The Turbine Building operator [Surry verify correct nomenclature] checks both condenser air ejector loop seal drain lines to condenser, and reports that one loop seal drain line is very hot to the touch, and the other is at normal temperature to the touch.

Based on the current conditions, which one of the following are the required actions, in accordance with 1-AP-14.00?

A. Isolate ONLY the hot loop seal drain line, verify the condenser hoggers are in service, and secure ONLY the set of air ejectors associated with the hot loop seal drain line.

B. Isolate ONLY the hot loop seal drain line, if Air Ejector flows are NOT normal leave the drain valve closed for approximately 5 minutes, then reopen the hot loop seal drain isolation valve.

C. Isolate BOTH loop seal drain lines, verify the condenser hoggers are in service, and secure ONLY the set of air ejectors associated with the hot loop seal drain line.

D. Isolate BOTH loop seal drain lines, if Air Ejector flows are NOT normal leave both drain valves closed for approximately 5 minutes, then reopen both loop seal drain isolation valves.

K/A Loss of Condenser Vacuum 051AG2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects.

(CFR 41.10 / 43.5 / 45.13) (RO - 3.8)

K/A Match Analysis This question matches the K/A statement by requiring the applicant to correctly apply knowledge of the AP for loss of condenser vacuum when directing local auxiliary operator tasks, given an operationally plausible situation. An auxiliary operator may not have the procedure in hand when performing local tasks during an emergency, and may rely on guidance provided from the control room.

Answer Choice Analysis A. INCORRECT. All parts of this distractor are incorrect. The correct actions, as specified in Attachment 3 of 1-AP-14.00, are to close both drain valves, wait about five minutes, and then reopen the drain valves. The candidate who does not understand or remember the correct actions from the AP may believe these actions to be correct, because it is plausible that the hot drain line is the one that is malfunctioning. I included the statement about the hoggers in operation for securing an air ejector to be a plausible action during a loss of condenser vacuum casualty.

B. INCORRECT. The actions in this distractor are correct; however, the procedure requires the actions be performed on both drain valves, not just the hot drain line.

C. INCORRECT. The initial set of actions is correct (to close both loop seal drain lines). However, the second set of actions (to secure a malfunctioning air ejector) is incorrect.

D. CORRECT. As detailed in step 4 of Attachment 3 of 1-AP-14.00, for this condition both drain valves are closed for approximately 5 minutes to refill the lines, and then both drain valves are re-opened.

Supporting References

1. Modified from question 051AG2.4.35 from the North Anna 2008-301 exam to support Surry references.

2. Surry Procedure 1-AP-14.00, "LOSS OF MAIN CONDENSER VACUUM," rev. 7.

Especially Attachment 3 step 4.

References Provided to Applicant

None.

Answer: D

23. 000000059AA1.01 1 Current conditions on Unit 1:

• A Large Break Loss of Coolant Accident has occurred.

• 1B train of Recirc Spray is the only available train and started automatically.

• Annunciator 1-RM-A8, RS/SW HX B ALERT / FAILURE is lit.

• The RS/SW HX B radiation monitor, 1-SW-RM-115, is indicating all EEEEEs.

• The following alarm lights are lit on the monitor:

• Warning

• High

• Range Based on the current conditions, which one of the following describes the response of the radiation monitors and an acceptable action in accordance with alarm response procedure 1-RM-A8?

A. 1-SW-RM-115 has failed high. Request HP sampling per the Offsite Dose Calculation Manual.

B. Current radiation levels are above the range of 1-SW-RM-115. Request HP sampling per the Offsite Dose Calculation Manual.

C. 1-SW-RM-115 has failed high. Secure 1B Recirc Spray train.

D. Current radiation levels are above the range of 1-SW-RM-115. Secure 1B Recirc Spray train.

K/A Accidental Liquid Radwaste Release Ability to operate and/or monitor the following as they apply to the Accidental Liquid Radwaste Release: Radioactive - liquid monitor.

(CFR: 41.7/45.5/45.6) (RO - 3.5)

K/A Match Analysis The RO applicant is required to monitor indications associated with the radiation

monitors on the Recirculation Spray heat exchangers, recognize that elevated readings exist, determine that an accidental liquid release is occurring and apply the correct actions to address the release.

Answer Choice Analysis NOTE: The question focuses on whether the applicant can determine from the indications of the radiation monitor whether the instrument has failed high or is detecting radiation levels outside the meters range.

A. INCORRECT. Plausible because per the note prior to step 1 of 1-RM-A8 the digital ratemeter with all EEEEEs displayed indicates a failure of the digital ratemeter.

However, all EEEEs with the warning, high and range alarm lights lit indicates that radiation levels are above the range of the meter. An upscale failure of the monitor would also cause the fail alarm light to light up. In addition, requesting HP sampling is the correct answer. Per the caution statement prior to step five of the ARP, the RS train should only be secured if a redundant train is available.

B. CORRECT. All EEEEs with the warning, high, and range alarm lights lit indicates that radiation levels are above the range of the meter. In addition, requesting HP sampling is the correct answer. The RNO actions for a RS heat exchanger that cannot be removed from service (See step 6 of the ARP) is either monitor the Discharge Tunnel radiation monitor or have HP conduct sampling per the ODCM.

C. INCORRECT. Plausible because per the note prior to step 1 of 1-RM-A8 the digital ratemeter with all EEEEEs displayed indicates a failure of the digital ratemeter.

However, all EEEEs with the warning, high and range alarm lights lit indicates that radiation levels are above the range of the meter. The second half of the response is plausible because on indications of a leak on an RS heat exchanger (high radiation reading), the ARP directs securing the RS train. However, per the caution statement prior to step five of the ARP, the RS train should only be secured if a redundant train is available.

D. INCORRECT. Plausible because the first half of the response is correct. The second half of the response is plausible because on indications of a leak on an RS heat exchanger (high radiation reading), the ARP directs securing the RS train. However, per the caution statement prior to step five of the ARP, the RS train should only be secured if a redundant train is available.

Supporting References

1. ND-93.5-LP-1, Pre-TMI Radiation Monitoring System, Rev. 10, Pgs. 5-7 and11-13

2. ND-91-LP-6, Recirculation Spray System, Rev. 13, pg. 12

3. 1-RM-A8, RS/SW HX B ALERT/FAILURE, Rev. 4

References Provided to Applicant None.

Answer: B

24. 000000067AA2.17 1 Unit 1 initial conditions:

Fire is reported in the "B" 4160v station service bus The SRO directs the bus to be de-energized Based on the above conditions, which one of the following states (1) the load lost when de-energizing the B 4160v station service bus and (2) the type of fire suppression system in the 4160V Station Service Switchgear room?

A. (1) 1-BC-P-1A "A" BC (2) Low Pressure CO2 B. (1) 1-SD-P-1B "B" HP (2) Low Pressure CO2 C. (1) 1-BC-P-1A "A" BC (2) High Pressure CO2 D. (1) 1-SD-P-1B "B" HP (2) High Pressure CO2 K/A Plant Fire On-Site. Ability to determine and interpret the following as they apply to the plant fire on Site: Systems that may be affected by the fire.

K/A Match Analysis Requires applicant to know how a fire on a specific switchgear will affect plant equipment.

Answer Choice Analysis A Correct: The 1-BC-P-1A "A" BC is powered from the "B" 4160v station service bus.

The fire suppression system used for the 4160V Station Service Switchgear rooms is Low pressure CO2.

B Incorrect: 1st part is incorrect because the 1-SD-P-1B "B" HP is powered from the C 4160v station service bus. 1st part is plausible because it is pump 1B B which by convention would be powered from the B switchgear. 2nd part is correct.

C Incorrect: 1st part is correct. 2nd part is incorrect because these switchgear rooms have LP CO2 systems. Plausible because they are CO2 systems.

D Incorrect: : 1st part is incorrect because the 1-SD-P-1B "B" HP is powered from the C 4160v station service bus. 1st part is plausible because it is pump 1B B which by convention would be powered from the B switchgear. 2nd part is incorrect because these switchgear rooms have LP CO2 systems. Plausible because they are CO2 systems.

Ref ND-90.2-LP-2 Obj: A References Provided to Applicant None.

Answer: A

25. 000000068AK2.07 1 Unit 1 Initial Conditions:

• A large electrical fire occurred in the Main Control Room.

• Operators implemented 0-FCA-1.00, LIMITING MCR FIRE.

• Both units Reactor and Turbine are tripped.

• All Reactor Coolant Pumps (RCPs) are placed in Pull-to-Lock (PTL).

• During the confusion, the Unit 1 operator inadvertently placed ALL Unit 1 Charging Pumps in PTL.

• A Charging Pump is running on Unit 2.

Current Conditions:

• You were dispatched to the #1 EDG with 0-FCA-12.00, EMERGENCY DIESEL GENERATOR OPERATION.

• Upon arrival at #1 EDG, you report that the #1 EDG is NOT running.

• The Senior Reactor Operator (SRO) informs you that the station has NOT lost normal offsite power.

Based on the current conditions, which one of the following is (1) the required action for RCP seal injection, in accordance with 0-FCA-1.00, AND (2) the required action(s) for operation of the #1 EDG, in accordance with 0-FCA-12.00?

A. (1) RCP seals do NOT have to be isolated until 30 minutes have elapsed from the loss of seal injection flow.

(2) Verify switch lineup at #1 EDG for normal automatic operation, but do NOT transfer #1 EDG control to the local panel.

B. (1) RCP seals do NOT have to be isolated until 30 minutes have elapsed from the loss of seal injection flow.

(2) Transfer #1 EDG control to the local panel.

C. (1) IF RCP Seal Injection has been lost, THEN RCP seals must be isolated and remain isolated until the RCS is cooled down to less than 200 °F.

(2) Transfer #1 EDG control to the local panel.

D. (1) IF RCP Seal Injection has been lost, THEN RCP seals must be isolated and remain isolated until the RCS is cooled down to less than 200 °F.

(2) Verify switch lineup at #1 EDG for normal automatic operation, but do NOT transfer #1 EDG control to the local panel.

K/A 068 Control Room Evacuation 068AK2.07 Knowledge of the interrelations between the Control Room Evacuation and the following: ED/G.

(CFR 41.7 / 45.7) (RO - 3.3)

K/A Match Analysis This question matches the K/A statement by requiring the RO applicant to demonstrate knowledge of the overall mitigating strategy for #1 EDG operation during a Control Room Evacuation casualty when offsite power is NOT lost to the station. In the other part of the question, the RO applicant will demonstrate knowledge of RCP seal isolation during a Control Room Evacuation casualty, given a plausible operationally valid situation (loss of seal injection flow coincident with Control Room Evacuation).

Answer Choice Analysis A. INCORRECT. Part (1) is incorrect. A CAUTION statement before step 20 of 0-FCA-1.00, LIMITING MCR FIRE, (CHECK CHARGING PUMPS - AT LEAST ONE RUNNING ON EACH UNIT) clearly states: If RCP Seal Injection has been lost, RCP seals must be isolated and remain isolated until the RCS is cooled down to less than 200 °F. Part (1) is plausible because 30 minutes is a realistic time frame to RCP seal failure. Part (2) is also incorrect. Given the operational situation, the operator will use 0-FCA-12.00 to verify the EDG switch lineup, then transfer EDG control to the local panel, then verify offsite power available and breaker 15H8 closed, and then transition to step 34 to determine if #3 EDG needs to be fast started. Part (2) is plausible because the question stem does not mention any effects on the #1 EDG automatic

functionsso it is plausible that the EDG will auto-start and load as needed if offsite power was lost. However, the procedure 0-FCA-12.00 directs transferring control of the #1 EDG to local in every circumstance.

B. INCORRECT. Part (1) is incorrect, part (2) is correct. See analysis of A. above.

C. CORRECT. Part (1) is correct, part (2) is correct. See analysis of A. above.

D. INCORRECT. Part (1) is correct, part (2) is incorrect. See analysis of A. above.

Supporting References

1. Surry Procedure 0-FCA-1.00, LIMITING MCR FIRE, rev. 42. Especially step 20 and its CAUTION statement.

2. Surry Procedure 0-FCA-12.00, EMERGENCY DIESEL GENERATOR OPERATION, rev. 14. Procedural flow path for operation of #1 EDG for a Control Room Evacuation without a loss of offsite power.

References Provided to Applicant None.

Answer: C

26. 000000WE08EK1.1 1 Which one of the following correctly states the SI reduction criteria for 1-FR-P.1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITION (1) for subcooling margin and (2) the RCS inventory with all RCPs running?

A. (1) > 30 F [85 F]

(2) > 82% Dynamic Range RVLIS B. (1) > 30 F [85 F]

(2) > 22% [50%] PZRZ level C. (1) > 80 F [135 F]

(2) > 82% Dynamic Range RVLIS D. (1) > 80 F [135 F]

(2) > 22% [50%] PZRZ level K/A

RCS Overcooling Knowledge of operational implications of the following concepts as they apply to the (Pressurizer Thermal Shock): Components, capacity, and function of emergency systems.

K/A Match Analysis To arrive at the correct answer, the applicant must know the SI reduction criteria from P.1. This is testing knowledge of components, systems, and their function as it relates to PTS.

Answer Choice Analysis A. INCORRECT. The second part is correct. The first part is plausible because this is the criteria used in E-2.

B. INCORRECT. Both parts are plausible because this is the criteria used in E-2.

C. CORRECT. These criteria are spelled out in step 6 of P.1.

D. INCORRECT. The first part is correct. The second part is plausible because this is the criteria used in E-2.

Supporting References

1. 1-E-2, FAULTED STEAM GENERATOR ISOLATION, Rev. 17.

2. 1-FR-P.1, RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITION, Rev. 17.

References Provided to Applicant None.

Answer: C

27. 000000WE10EK1.2 1 Current conditions:

- Unit 1 experienced a reactor trip and a loss of offsite power.

ES-0.4, Natural Circulation Cooldown with Steam Void in Rx Vessel (W/O RVLIS),

is in progress.

During RCS depressurization with Pressurizer level above 90% per 1-ES-0.4, which one of the following

(1) explains why Pressurizer heaters are energized and (2) identifies the parameter monitored to determine when the action is complete.

A. 1) Promotes cooling in the upper head region of the vessel.

2) RCS subcooling.

B. 1) Establish conditions to allow restart of a RCP when power is restored.

2) RCS subcooling.

C. 1) Promotes cooling in the upper head region of the vessel.

2) RCS pressure.

D. 1) Establish conditions to allow restart of RCP when power is restored.

2) RCS pressure.

K/A Natural Circulation Knowledge of the operational implications of the following concepts as they apply to the (Natural Circulation with Steam Void in Vessel with/without RVLIS): Normal, abnormal, emergency operating procedures associated with (Natural Circulation with Steam Void in Vessel with/without RVLIS)

(CFR: 41.8/41.10/45.3) (RO - 3.5)

K/A Match Analysis The RO applicant is required to know the operational reason for the action requiring energizing the pressurizer heaters when pressurizer level is greater than 90%.

Answer Choice Analysis A. INCORRECT. Plausible because the first part of the choice is correct. The second part is plausible because RCS subcooling is a parameter that is monitored at various steps throughout 1-ES-0.4, specifically as part of the actions associated with trying to start a RCP (Step 1).

B. INCORRECT. Plausible because attempting to start a RCP is Step 1 of 1-ES-0.4.

In addition, this action also has Pressurizer level as a condition of completion ( >68% vs

>90%) and addresses the energizing of the Pressurizer heaters as needed. The second part is plausible because RCS subcooling is a parameter that is monitored at various steps throughout 1-ES-0.4, specifically as part of the actions associated with trying to start a RCP (Step 1).

C. CORRECT. The RNO actions Step 10 of 1-ES-0.4 directs increasing RCS pressure by 100 psig using Pressurizer heaters if Pressurizer level is above 90%. The reason for this action is to promote heat removal from the upper head of the vessel by partially or completely collapsing any void that has built up in this region.

D. INCORRECT. Plausible because attempting to start a RCP is Step 1 of 1-ES-0.4.

In addition, this action also has Pressurizer level as a condition of completion ( >68% vs

>90%) and addresses the energizing of the Pressurizer heaters as needed. The second half of the question is correct - it is the parameter that is monitored to determine when to shut off the heaters.

Supporting References

1. Surry lesson plan ND-95.3-LP-55, "Emergency Response Guidelines ES-0.4, NC Cooldown with Steam Void in Rx Vessel (without RVLIS)," rev. 10, Obj B, pg. 19.

2. 1- ES-0.4, NC Cooldown with Steam Void in Rx Vessel (without RVLIS), rev. 12, pg. 5.

3. This question is modified from Sequoyah 2009-301 (Q #26). Modified to a two-part question stem requiring the reason PZR heater energizing during RCS cooldown and depressurization and the parameters used to determine when use of PZR heaters is complete.

References Provided to Applicant None.

Answer: C

28. 000003K4.07 1 Unit 1 Initial Conditions:

• 100% Power.

Current Conditions:

• No. 1 Seal Leakoff on 1-RC-P-1B, the B Reactor Coolant Pump (RCP),

unexpectedly rises to a value of 6.5 gpm and stable.

• Operators enter 1-AP-9.00, RCP ABNORMAL CONDITIONS.

• All temperatures, and all other monitored parameters on 1-RC-P-1B are within normal operating limits and are stable.

Based on the current conditions, which one of the following:

(3) is the required action, in accordance with 1-AP-9.00, AND

(4) describes the No. 2 seal design for 1-RC-P-1B during normal operation?

A. (1) Manually trip the Reactor, and secure the B RCP within 5 minutes.

(2) The No. 2 seal is a film-riding seal consisting of a runner which rotates with the shaft and a non-rotating seal ring attached to the lower seal housing.

B. (1) Manually trip the Reactor, and secure the B RCP within 5 minutes.

(2) The No. 2 seal is a rubbing-face seal comprised of a chrome carbide seal ring and a chrome-carbide runner that is fixed on the pump shaft.

C. (1) Shut down the plant, and secure the B RCP within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

(2) The No. 2 seal is a rubbing-face seal comprised of a chrome carbide seal ring and a chrome-carbide runner that is fixed on the pump shaft.

D. (1) Shut down the plant, and secure the B RCP within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

(2) The No. 2 seal is a film-riding seal consisting of a runner which rotates with the shaft and a non-rotating seal ring attached to the lower seal housing.

K/A Reactor Coolant Pump 003K4.07 Knowledge of RCPS design feature(s) and/or interlock(s) which provide for the following: Minimizing RCS leakage (mechanical seals).

(CFR 41.7 ) (RO - 3.2)

K/A Match Analysis This question matches the K/A statement by requiring the applicant to correctly remember the basic design of the No. 2 RCP seal (vs. the design of the No. 1 seal).

The question further recalls the applicant to determine the correct required action, given an operationally valid scenario involving the failure of a No. 1 RCP seal. This part of the question indirectly tests knowledge of the No. 2 RCP seal; whether it is capable of minimizing leakage during a controlled plant shutdown, or whether an immediate trip and RCP shutdown is required.

Answer Choice Analysis A. INCORRECT. The first part of the distractor is incorrect. This part is plausible because it is the action to take when seal leakoff is greater than 8 gpm and temperatures are increasing. However, the stem of the question states that all parameters are stable. The second part of the distractor is also incorrect, but plausible, because it describes the No. 1 RCP seal design exactly as the lesson plan describes it.

B. INCORRECT. The first part of the distractor is incorrect as detailed in A. above.

The second part of the distractor is correct; the description of the No. 2 RCP seal design is taken directly from the RCP lesson plan.

C. CORRECT. As detailed in the CAUTION statement before step 39 of 1-AP-9.00, with a high No. 1 seal leakoff (>8gpm) condition and no other parameters rising, the required action is to shut down the plant and secure the affected RCP within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

The second part distractor is also correct, as detailed above.

D. INCORRECT. See above descriptions.

Supporting References

1. Surry Procedure 1-AP-9.00, RCP ABNORMAL CONDITIONS, rev. 28. Especially CAUTION statement before step 39.

2. Surry Lesson Plan ND-88.1-LP-6, REACTOR COOLANT PUMPS, rev. 20 dtd 12/16/09.

References Provided to Applicant None.

Answer: C

29. 000004A2.25 1 Current conditions:

• The latest shutdown margin calculation indicates the shutdown margin has decreased to 1.7% delta-K/K.

• RCP A is in operation and preparations are underway to shutdown the pump.

• RHR system is in operation.

• RCS pressure has been reduced to 300 psig.

• RCS temperature is 210°F.

Based on the current conditions, which one of the following completes the statements below?

A tube leak exists on the (1) . Shutdown Margin as defined by Technical Specifications is (2) .

(1) (2)

A. Seal water return heat exchanger. met.

B. Seal water return heat exchanger. NOT met.

C. In-service RHR heat exchanger. met.

D. In-service RHR heat exchanger. NOT met.

K/A Chemical and Volume Control Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: uncontrolled boration or dilution.

(CFR: 41.5/43.5/45.3/45.5) (RO - 3.8)

Question History: Modified Bank (Sequoyah 2009-302, Q#30). Changed plant conditions from operation to Cold Shutdown and its effects on shutdown margin versus control rod movement.

LOK: C/A LOD: 3 K/A Match Analysis The RO applicant is required to identify a possible cause of an uncontrolled dilution via the CVCS and determine the actions required when shutdown margin has been reduced below 1.77% delta-K/K.

Answer Choice Analysis NOTE: RCS pressure is maintained at approximately 300 psig when establishing RHR in shutdown cooling. An RCS temperature of 210°F is consistent with the current plant conditions. GOP-2.5, Unit Cooldown - 351°F to less than 205°F, doesnt provide a specific temperature band, but directs placing RHR in service once RCS pressure is approximately 300 psig.

A. INCORRECT. Plausible because the first part of the answer is correct. Plausible because at Cold Shutdown (i.e. RCS temperature less than 200°F) the shutdown margin limit is 1.0% delta-K/K. However, in the Intermediate Shutdown mode (i.e. RCS temperature between 200°F and 547°F), the minimum shutdown margin is 1.77%

delta-K/K.

B. CORRECT. A leak on the Seal Water Return heat exchanger would allow Component Cooling (CC) water to enter the CVCS and cause an uncontrolled dilution of the RCS. The current shutdown margin of 1.7% delta-K/K does not meet the shutdown margin of 1.77% delta-K/K required for Intermediate Shutdown.

C. INCORRECT. Plausible because a CC water cools the RHR heat exchanger.

However, at the current RCS pressure (300 psig), a tube leak would result in flow from the RCS rather than into the RCS. In the Intermediate Shutdown mode, the minimum shutdown margin is 1.77% delta-K/K. Plausible because at Cold Shutdown (i.e. RCS temperature less than 200°F) the shutdown margin limit is 1.0% delta-K/K.

D. INCORRECT. Plausible because a CC water cools the RHR heat exchanger.

However, at the current RCS pressure (300 psig), a tube leak would result in flow from the RCS rather than into the RCS. The second half of the choice is correct.

Supporting References

1. Surry lesson plan ND-95.1-LP-10, "Dilution Accident," rev. 8, Obj B., pg. 6.

2. Surry lesson plan ND-88.2-LP-2, Operations of RHR System, Obj B., pp. 7-13.

3. T.S. 1.0 Definitions, Reactor Operations, pg. T.S 1.0-1, Amdnt 203

4. T.S. 3.12.G, Shutdown Margin, Amdnts 265 and 264.

References Provided to Applicant None Answer: B

30. 000004K6.36 1 Current conditions:

• Unit 1 is 100% power.

• The following alarms have been received on the 1D annunciator panel.

• F1 - VCT HI TEMP

• G3 - DEMIN INL DIVERT HI TEMP

• 1-CH-FI-1150, Letdown Line Flow, is fluctuating erratically.

Which ONE of the following failures could account for the above conditions?

A. The instrument air line to 1-CH-PCV-1145, Letdown Line Pressure Control, has ruptured.

B. 1-CH-PT-1145, Letdown Line Pressure Transmitter, has failed downscale.

C. The instrument air line to 1-CC-TCV-103, NRHX Outlet Temperature Control, has ruptured.

D. 1-CH-TE-1144, NRHX Outlet Temperature Transmitter, to 1-CC-TCV-103 has failed

upscale.

K/A Chemical and Volume Control Knowledge of the effect of a loss or malfunction on the following CVCS components:

Letdown pressure control to prevent RCS coolant from flashing to steam in letdown piping.

(CFR: 41.7/45.7) (RO - 2.9)

Question History: New.

K/A Match Analysis The RO applicant is required to evaluate the plant conditions and determine that the backpressure valve (1-CH-PCV-1145) has failed open (PCV-1145 fails open on loss of air).

Answer Choice Analysis A. CORRECT. A loss of instrument air to 1-CH-PCV-1145 will cause the valve to open.

The resulting transient will allow the fluid to flash to steam and create a downstream pressure and temperature transient that will cause flow to be diverted around the ion exchangers and lift RV-1209 which will route high temperature water to the VCT.

B. INCORRECT. Plausible because a failure of this pressure transmitter could cause 1-CH-PCV-1145 to open. However, since the transmitter senses pressure upstream of the valve a downscale failure would cause the valve to close rather than open.

C. INCORRECT. Plausible because closure of the 1-CC-TCV-103 would result in the same indications provided in the stem of the question. However, 1-CC-TCV-103 fails open on a loss of instrument air.

D. INCORRECT. Plausible because a failure of the temperature transmitter could cause 1-CC-TCV-103 to close. However, an upscale failure of this transmitter would cause 1-CC-TCV-103 to open.

Supporting References

1. Surry lesson plan ND-88.3-LP-2, "Charging and Letdown," rev. 15, Obj C, p. 15.

2. 1D-F1, VCT HI TEMP, Rev.0

3. 1D-G3, DEMIN INL DIVERT HI TEMP, Rev. 1

4. Dwg # 11448-FM-072C, Sh. 4 of 5, Rev. 28

References Provided to Applicant None Answer: A

31. 000005A4.01 1 Unit 1 Initial Conditions:

• The plant is solid.

• RCS temperature is being controlled in auto with the A RHR pump running. The B RHR pump is not running.

• RCS pressure is being controlled with the PCV-145 controller in MANUAL and the FCV-122 controller in MANUAL.

Current conditions:

• The A RHR pump just tripped.

Based on the current conditions, which one of the following (1) describes the response of RCS pressure if NO operator actions are taken, AND (2) the action the operator is required to take to restore RCS pressure to its value before the A RHR pump tripped?

A. (1) RCS pressure will initially increase, and then steadily decrease (2) RAISE demand on PCV-145 controller to raise RCS pressure B. (1) RCS pressure will initially increase, and then steadily decrease (2) LOWER demand on PCV-145 controller to raise RCS pressure C. (1) RCS pressure will initially decrease, and then steadily increase (2) RAISE demand on FCV-122 controller to reduce RCS pressure D. (1) RCS pressure will initially decrease, and then steadily increase (2) LOWER demand on FCV-122 controller to reduce RCS pressure K/A 005 Residual Heat Removal 005A4.01 Ability to manually operate and/or monitor in the control room: Controls and indication for RHR pumps.

(CFR 41.7 / 45.5 to 45.8) (RO - 3.6)

K/A Match Analysis

This question matches the K/A statement by requiring the RO applicant to correctly apply knowledge of the integrated plant response to an RHR pump trip with the plant in an operationally valid situation (solid condition). The RO applicant is required to understand how monitoring the RHR pump status relates to expected plant response, as well as how to operate controls in manual to mitigate the effects of the RHR pump trip on RCS pressure.

Answer Choice Analysis A. INCORRECT. RCS pressure will initially decrease and then steadily increase, raising demand on PCV-145 controller will throttle open the valve, which will lower RCS pressure.

B. INCORRECT. RCS pressure will initially decrease and then steadily increase, and lowering demand on PCV-145 controller will throttle closed the valve, which will lower RCS pressure.

C. INCORRECT. RCS pressure will initially decrease when the RHR pump trips, then steadily increase. Opening FCV-122 controller would continue to raise pressure above the pre-RHR pump trip value.

D. CORRECT. RCS pressure will initially decrease and then steadily increase, and lowering demand on the FCV-122 controller will lower RCS pressure.

Supporting References

1. Modified from VC Summer 2004-301 Question 005A4.01.

2. Surry Lesson Plan ND-88.2-LP-2, Operation of Residual Heat Removal System, rev. 16 dtd. 11/21/06. Especially section 3.1, Solid Plant Ops.

References Provided to Applicant None.

Answer: D

32. 000006K1.07 1 Which one of the following correctly states the minimum parameter(s) that are required for closure of the Feedwater Regulating Valves and Feedwater Regulating Bypass Valves?

A. Steam header pressure is 120 psig above steam line pressure for at least two steam lines.

B. Pressurizer pressure is less than 1885 psig.

AND Median Tave less than 554° F.

C. Pressurizer pressure is less than 1780 psig.

D. Pressurizer pressure is less than 1885 psig.

K/A Emergency Core Cooling Knowledge of the physical connections and/or cause-effect relationships between the ECCS and the following systems: MFW System.

(CFR: 41.2 to 41.9 /45.7 to 45.8) (RO - 2.9)

K/A Match Analysis The RO applicant is required to recognize the parameters and protection logic trends that will lead to a Safety Injection actuation and cause a Main Feedwater Isolation.

Answer Choice Analysis A. INCORRECT. Plausible because a differential pressure of more than 120 psig between the main steam header and the main steam lines is an input for an automatic SI actuation. The response also contains the correct coincidence for the protection channels. However, the differential pressure only needs to be sensed on one steam line not two.

B. INCORRECT. Plausible because with pressurizer pressure below 1885 psig a Reactor Protection trip will occur (i.e., reactor trip breakers open). This coupled with Median Tave below 554°F will cause a closure of the Main Feed Reg Valves but not a complete MFW isolation.

C. CORRECT. When pressurizer pressure reaches 1780 psig an SI actuation will occur which will in turn cause a MFW isolation.

D. INCORRECT. Plausible because steam pressure below 525 psig on two steam lines is an input for an SI actuation. However, this condition must be coupled with high steam flow (109%) on two steam generators to generate an SI actuation and subsequent MFW isolation.

Supporting References

1. Surry lesson plan ND-91-LP-3, "SI System Operations," rev. 20, Objective A and D, pp 3-7, 13-14.

2. Surry lesson plan ND-89.3-LP-3, Main Feedwater System, Rev. 21, Objective B, pgs. 10-11.

References Provided to Applicant None.

Answer: C

33. 000007A2.05 1 The following Unit 1 conditions exist:

Pressurizer PORV leakage occurs PRT pressure is 12 psig PRT gas samples indicate Xe-133 activity at 8x10-2 micro-curies/ml Operators have been directed to vent the PRT Which one of the following describes where the PRT is required to be vented in accordance with 1-OP-RC-011, Pressurizer Relief Tank Operations?

A. Vent Vent System B. Overhead Gas System C. Through the Sample System to the Process Vent System D. Process Vent System directly (I.E., not through the Sample System)

K/A Pressurizer Relief / Quench Tank A2.05 Ability to (a) predict the impacts of the following malfunctions or operations on the PRTS: and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Exceeding PRT high-pressure limits.

K/A Match Analysis The PRT normal pressure band is 2-4 psig and the high pressure alarm annunciates at 10 psig. Pressure can be vented to clear the alarm and get back within the normal operating band. The K/A requires the applicant to use knowledge of procedures to address exceeding a PRT high pressure limit. This question does require the applicant to have procedural knowledge of how to vent the PRT under a high pressure condition

with activity in the PRT.

Answer Choice Analysis A. Incorrect: due to the CAUTION statements in the OP. Plausible because the Vent Vent System is one procedurally directed option for venting the PRT under different conditions.

B. Correct: With Xe-133 activity above 5x10-2 micro-curie/ml, the CAUTION statements at the top of the PRT venting sections of 1-OP-RC-011 require the PRT to be vented to the Overhead Gas System.

C. Incorrect: due to the CAUTION statements in the OP. Plausible because the Sample System via the Process Vent System is one procedurally directed option for venting the PRT under different conditions.

D. Incorrect: due to the CAUTION statements in the OP. Plausible because the Process Vent System is one procedurally directed option for venting the PRT under different conditions.

Supporting References 1-OP-RC-011, Pressurizer Relief Tank Operations, Rev. 24.

Memory Level, RO, LOD=2, NEW References Provided to Applicant none Answer: B

34. 000008A3.05 1 Unit 1 Initial Conditions:

• Power = 100%.

Current Conditions:

• Reactor Coolant System (RCS) Pressure is 900 psig and slowly lowering.

• Pressurizer level is offscale low.

• Containment pressure reached a maximum at 19 psia and is now 18 psia and slowly lowering.

• Steam Generator pressures and levels are equal and stable.

Based on the current conditions, which one of the following is the Component Cooling

(CC) valve lineup? (consider that ONLY automatic actuations occurred, no manual re-positioning of valves)

CC-TV-105A, B, C Reactor Coolant Pump (RCP) Cooler Return Valves CC-TV-110A, B, C Containment Air Recirc CC Return Valves

-105A, B, C -110A, B, C A. CLOSED CLOSED B. OPEN CLOSED C. CLOSED OPEN D. OPEN OPEN K/A 008 Component Cooling Water 008A3.05 Ability to monitor automatic operation of the CCWS, including: Control of the electrically operated, automatic isolation valves in the CCWS.

(CFR 41.7 / 45.5) (RO - 3.0)

K/A Match Analysis This question matches the K/A statement by requiring the RO applicant to apply knowledge of the automatic isolation valve logic for important CCW valves, given an operationally valid scenario of an SI signal, Hi-CLS signal, but NO HI-HI-CLS signal.

Answer Choice Analysis A. INCORRECT. The question stem states that RCS Pressure is 900 psig and lowering; therefore, an SI signal was actuated. Also, with a peak containment pressure of 19 psia, a HI-CLS signal would be actuated (HI-CLS setpoint is 17.7 psia), but a HI-HI-CLS signal would NOT have actuated (HI-HI-CLS setpoint of 23 psia). The Surry lesson plan for CCW states that the -109 valves auto close on SI, the -110 valves auto close on HI-HI CLS, and the -105 valves also auto close only on HI-HI CLS. Therefore, the correct alignment is the -109 valves CLOSED, the -110 valves OPEN, and the -105 valves OPEN. All other combinations are plausible misconceptions of the correct alignment.

B. INCORRECT. See analysis of A above.

C. CORRECT. See analysis of A above.

D. INCORRECT. See analysis of A above.

Supporting References

1. Surry Lesson Plan ND-88.5-LP-1, COMPONENT COOLING WATER, rev. 23.

2. Surry Lesson Plan ND-88.4-LP-2, CONTAINMENT VESEEL, rev. 12. Page 15 gives the logic setpoints for the CLS signals.

3. Extensively modified from a DC Cook 2001 008A3.05 ILO exam question to apply to Surry.

References Provided to Applicant None.

Answer: D

35. 000008K4.09 1 Unit 1 Initial Conditions:

• Power = 100%.

• Component Cooling (CC) Pump 1-CC-P-1B is RUNNING.

• An electrical fault caused a loss of normal offsite power to the 1H Emergency Bus.

• Operators took action in accordance with 1-AP-10.07, LOSS OF UNIT 1 POWER, and have re-energized the 15H10 stub bus from the #1 EDG.

• Operators have completed 1-AP-10.07 and are at the 'when/then' step to initiate 0-AP-10.08, STATION POWER RESTORATION, once the cause of power loss is corrected.

Current Conditions:

• Time = 0800. A Large Break LOCA occurs.

• Time = 0802. Containment pressure peaks at 28 psia.

• Time = 0810. 1-CC-P-1B trips.

Which one of the following completes the below statements?

(1) The standby CC pump low discharge header pressure auto start setpoint is (2) Based on the current conditions, when CC discharge pressure reaches the low

discharge header pressure auto start setpoint, CC Pump 1-CC-P-1A will A. (1) 55 psig (2) automatically start on low discharge header pressure B. (1) 55 psig (2) NOT automatically start on low discharge header pressure C. (1) 75 psig (2) automatically start on low discharge header pressure D. (1) 75 psig (2) NOT automatically start on low discharge header pressure K/A 008 Component Cooling Water 008K4.09 Knowledge of the CCWS design feature(s) and/or interlock(s) which provide for the following: The standby feature for the CCW pumps.

(CFR 41.7 ) (RO - 2.7)

K/A Match Analysis This question matches the K/A statement by requiring the RO applicant to demonstrate knowledge of the correct auto-start setpoint for low CCW discharge pressure, as well as demonstrate that there is no interlock preventing the standby CCW pump from auto-starting in the presence of a HI-HI-CLS signal, although the procedures are full of CAUTIONS stating that this is not a desired condition.

Answer Choice Analysis A. CORRECT. The Surry lesson plan for Component Cooling Water (CC) states:

The standby pump will auto start on a low discharge header pressure of 55 psig.

Working through the 1-AP-10.07 procedure for a loss of normal offsite power to the 1H bus, the standby CC pump (1-CC-P-1A) will be placed in PTL, the stub bus energized from the diesel, and then the CC pump switch will be returned to AUTO-AFTER-STOP.

There is no electrical interlock preventing the standby CC pump from auto starting on low discharge header pressure; however, there is a CAUTION statement before step 47 in 1-AP-10.07 that states: When the EDG is the only source of power to an Emergency Bus, the associated Component Cooling Pump should NOT be in service if a HI-HI CLS is in progress. The question stem sets up a situation where HI-HI CLS has actuated

(> 23 psia in containment).

B. INCORRECT. Part (1) is correct, part (2) incorrect. See analysis of A above.

C. INCORRECT. Part (1) is incorrect, part (2) is correct. See analysis of A above.

D. INCORRECT. Parts (1) and (2) both incorrect. See analysis of A above.

Supporting References

1. Surry Lesson Plan ND-88.5-LP-1, COMPONENT COOLING SYSTEM, rev. 23.

2. Surry Procedure 1-AP-10.07, LOSS OF UNIT 1 POWER, rev. 54. Especially NOTE before step 47.

2. Surry Lesson Plan ND-90.3-LP-7, STATION SERVICE AND EMERGENCY DISTRIBUTION PROTECTION AND CONTROL, rev. 23.

References Provided to Applicant None.

Answer: A

36. 000010K5.02 1 Unit 1 Initial Conditions:

• 100% Power.

• Pressurizer Relief Tank (PRT) pressure is 2.2 psig.

• A transient causes a Pressurizer (PZR) PORV to open.

• The PZR PORV will not reseat, and operators are unable to close the associated PORV block valve.

• Maximum PZR pressure was 2340 psig and has continually DECREASED from that value.

Current conditions:

• Containment pressure is 10.62 psia.

• PRT Pressure is 90 psig and INCREASING at 10 psig/min.

• PZR Pressure is 1550 psig and DECREASING at 5 psig/min.

Based on the current conditions, and assuming a completely ideal thermodynamic process, which one of the following is (1) the expected temperature trend observed on TE-463, PZR PORV downstream temperature instrument, from the maximum observed PZR pressure to the current

conditions AND (2) the expected temperature value read on TE-463 two (2) minutes after the current conditions, considering no operator actions and all given trends continuing for the two (2) minute interval?

Expected TE-463 temp Expected TE-463 temp trend two minutes later A. INCREASING GREATER THAN current temperature on TE-463 B. INCREASING LESS THAN current temperature on TE-463 C. DECREASING GREATER THAN current temperature on TE-463 D. DECREASING LESS THAN current temperature on TE-463 K/A Pressurizer Pressure Control 010K5.02 Knowledge of the operational implications of the following concepts as they apply to the PZR PCS: Constant enthalpy expansion through a valve.

(CFR 41.5 / 45.7 ) (RO - 2.6)

K/A Match Analysis The question author wanted to examine the applicants fundamental understanding of the thermodynamic process, instead of a GFES-style temperature-lookup-with-the-steam tables question. The question matches the K/A statement by requiring the applicants to correctly understand how the Mollier diagram

describes the constant enthalpy (throttling) process for an operationally valid situation.

To arrive at the correct answer, the applicant must either use the Mollier diagram to determine that downstream tailpipe temperatures would be increasing, until the point at which the PRT rupture disc fails (at 100 psig), at which time the PZR PORV will be exhausting to a containment volume that is at an initially subatmospheric pressure.

Answer Choice Analysis A. INCORRECT. Part (1) of this distractor is correct; as the PORV discharges to the saturated PRT system and PRT pressure increases, the downstream/tailpipe temperature is expected to increase (essentially following the saturation temperature for the given PRT pressure). Part (2) of this distractor is incorrect; two minutes after the current conditions, the PRT rupture disc will fail, and the PORV will be exhausting to the containment volume which is initially at a vacuum. Because the rupture disc fails, the expected TE-463 temperature will be less than the reading at current conditions. It is plausible for an applicant to believe that the pressure would continue to increase if the applicant does not recognize that the PRT will rupture at 100 psig.

B. CORRECT. See discussion for A. above. Plotting a few points will also get the correct answer:

-at PZR P=2340 psig, PRT P=2.2 psig, TE-463 ~ 220 °F

-at PZR P=2000 psig, PRT P=4.2 psig, TE-463 ~ 226 °F

-at PZR P=1700 psig, PRT P=20 psig, TE-463 ~ 258 °F

-at PZR P=1550 psig, PRT P=90 psig, TE-463 ~ 331 °F [current conditions]

-at PZR P=1450 psig, PRT P=Containment P ~11 psia, TE-463 ~ 198 °F [conditions 2 min after current conditions].

C. INCORRECT. Part (1) of this distractor is incorrect. This part of the distractor is plausible if the applicant believes that the downstream temperature will follow the PZR steam space temperatures, which will be decreasing as the pressurizer depressurizes.

The second part of the distractor is plausible if the applicant mis-applies the Mollier diagram and believes that the peak saturation line pressure is 1500 psia (instead of the correct 500 psia). If the PZR was at 500 psia, as it continued to depressurize it could enter the superheated steam region.

D. INCORRECT. Part (1) of this distractor is incorrect, as discussed above in the analysis of distractor C. Part (2) of this distractor is correct; in this case, plausibility is actually enhanced by the part (1) misconception, if the applicant simply believes that the current trends will continue for a two minute interval.

Supporting References

1. Surry Lesson Plan ND-88.1-LP-3-DRR, Pressurizer and Power Relief, rev. 16 dtd 07/07/08. Designates TE-463 as downstream PORV temperature instrument, PRT rupture disc set for 100 psig.

2. Steam Tables.

References Provided to Applicant Steam Tables.

Answer: B

37. 000012K2.01 1 Unit 1 Initial Conditions:

• 5% Power.

• The 'A' reactor trip breaker and the 'B' reactor trip bypass breaker are CLOSED.

Current conditions:

• The 'A' DC bus loses power/de-energizes.

Based on the current conditions, which one of the following is the automatic response of the Reactor Protection System?

'A' Reactor Trip Breaker 'B' Reactor Trip Bypass Breaker A. OPEN CLOSED B. CLOSED OPEN C. CLOSED CLOSED D. OPEN OPEN K/A 012 Reactor Protection 012K2.01 Knowledge of bus power supplies to the following: RPS channels, components, and interconnections.

(CFR 41.7) (RO - 3.3)

K/A Match Analysis The question matches the K/A by straightforwardly examining the RO applicants knowledge of the bus power supplies to RPS channels, with a slightly abnormal

situation (a Rx trip bypass breaker closed) in order to raise the questions LOD.

Answer Choice Analysis A. INCORRECT. Plausible if applicant believes "B" train reactor trip bypass breaker UV coils are powered from "B" 125 VDC.

B. INCORRECT. Plausible if applicant believes "A" train reactor trip breaker UV coils are powered from "B" 125 VDC.

C. INCORRECT. Plausible if applicant does not recognize that 125 VDC powers reactor trip and bypass breakers.

D. CORRECT. The "A" 125V DC train supplies power to both the "A" reactor trip UV trip coils and the "B" reactor trip bypass breaker UV trip coils.

Supporting References

1. Surry Lesson Plan ND-93.3-LP-10, Reactor Protection - General, rev. 7, dtd 12/04/2008.

2. Surry Lesson Plan ND-90.3-LP-6, 125 VDC Distribution, rev. 16, dtd 01/05/2010.

3. Slightly modified (not enough to not consider it BANK) from SURRY ILO exam 2006-301 012K2.01. This question was also based on Surry examination bank question number DC00012.

References Provided to Applicant None.

Answer: D

38. 000012K3.04 1 Unit 1 Initial Conditions:

• Time = 0300. Power = 100%, no equipment out of service.

• Pressurizer (PZR) Pressure channel 455 unexpectedly fails high.

Current conditions:

• Time = 0850 (same day). All required bistables for PZR Pressure channel 455 have been placed in the TRIP condition.

Based on the current conditions, which one of the following identifies the Reactor

Protection System (RPS) and Engineered Safety Function (ESF) actuation logic coincidence required, from the remaining in-service PZR Pressure protection channels, to initiate an automatic low PZR Pressure Reactor Trip and automatic Low PZR Pressure Safety Injection at time 0851?

Low PZR Pressure Low PZR Pressure Rx Trip Actuation Safety Injection Actuation Logic Coincidence Logic Coincidence A. 1/3 1/3 B. 1/2 1/2 C. 2/3 2/3 D. 1/3 1/2 K/A 012 Reactor Protection 012K3.04 Knowledge of the effect that a loss or malfunction of the RPS will have on the following: ESFAS.

(CFR 41.7 / 45.6) (RO - 3.8)

K/A Match Analysis This question matches the K/A statement by requiring the RO applicant to correctly apply knowledge of the actuation logic coincidence of the low pressurizer pressure reactor trip signal and the low pressurizer SI signal, given an operationally valid situation involving a malfunction of an RPS component, along with taking the Tech Spec required actions for that malfunction (placing the channel bistables to TRIP condition).

Answer Choice Analysis A. INCORRECT. Auto Rx Trip and auto SI is normally 2/3 for low pressurizer pressure. Channel 455 feeds both circuits. When a protection channel is removed from service, all bistables are tripped in all cases (except for the AUTO hi-hi CLS/Containment Spray actuation). Therefore, an auto SI will occur is either of the two remaining bistables trip, and an auto Rx trip will occur is either of the two remaining bistables trip. 1/3 and 2/3 are credible distractors because the applicant must know what state bistables will be in, and that some RPS functions (such as high PRNI flux) are 2 out of 4 logic under normal conditions.

B. CORRECT. See analysis of A above.

C. INCORRECT. See analysis of A above.

D. INCORRECT. See analysis of A above.

Supporting References

1. Modified from Harris 2006 Question 012K3.04.

2. Surry Procedure 0-DRP-004, PRECAUTIONS, LIMITATIONS AND SETPOINTS, rev. 63. Especially attachment 1, REACTOR PROTECTION SYSTEM SETPOINTS, 1.A.1. and 2.E.

3. Surry Technical Specifications 3.7, INSTRUMENTATION SYSTEMS, especially Table 3.7-1, Table 3.7-2, and associated ACTIONS.

4. Surry Lesson Plan ND-93.3-LP-10, Reactor Protection - General, rev. 7 dtd.

12/04/08.

References Provided to Applicant None.

Answer: B

39. 000013K5.02 1 On Unit 1 the 'A' steam line low pressure detector is out of service and Technical Specifications actions have been taken for that channel.

Which one of the following is now the minimum logic necessary for a valid, reliable Safety Injection (SI) signal to occur?

A. High steam flow on 1 of 2 channels on one of the three steam lines in coincidence with low pressure on either 'B' or 'C' steam line.

B. High steam flow on 1 of 2 channels on two of the three steam lines in coincidence with low pressure on either 'B' or 'C' steam line.

C. High steam flow on 2 of 2 channels on one of the three steam lines in coincidence with low pressure on both 'B' and 'C' steam lines.

D. High steam flow on 2 of 2 channels on two of the three steam lines in coincidence with low pressure on both 'B' and 'C' steam lines.

K/A Engineered Safety Features Actuation 0013 K5.02 Knowledge of the operational implications of the following concepts as they apply to ESFAS: Safety system logic and reliability.

(CFR 41.5 /45.7) (RO - 2.9)

K/A Match Analysis This question matches the K/A statement by requiring the applicants to correctly identify the logic of the SI actuation system (ie. system logic) with one steam line pressure detector out of service (i.e. reliability).

Answer Choice Analysis A. INCORRECT. Plausible because it only requires a high steam flow condition to be sensed on 1 of 2 channels, but it requires a high steam flow condition on 2 of 3 steam generators rather than only one steam generator. The logic for low steam pressure is correct. With one detector out of service its bistable would be placed in the tripped condition, so that a low steam line pressure on either 'B' or 'C' steam line will satisfy the 2 of 3 logic for low steam line pressure.

B. CORRECT. A high steam flow condition sensed on 1 of 2 channels on 2 of 3 steam generators coincident with low steam pressure on either 'B' or 'C' steam line will result in an SI actuation.

C. INCORRECT. Plausible because requiring 2 of 2 channels would ensure a channel faillure would not result in an input to the SI actuation logic. In addition, it requires a high steam flow condition on 2 of 3 steam generators rather than only one steam generator which ensures that a single channel failure would not result in an input to the SI actuation logic. The low steam line pressure on both 'B' and 'C' steam lines is plausible if the applicant doesn't recognize that the bistable for the failed detector is placed in the tripped condition as part of the Technical Specification actions.

D. INCORRECT. lausible because requiring 2 of 2 channels would ensure a channel faillure would not result in an input to the SI actuation logic. The response contains the correct logic for high steam flow (i.e. 2 of 3 steam generators). The low steam line pressure on both 'B' and 'C' steam lines is plausible if the applicant doesn't recognize that the bistable for the failed detector is placed in the tripped condition as part of the Technical Specification actions.

Supporting References

1. Surry lesson plan ND-91-LP-3, Safety Injection System Operations, Rev. 22, Obj. A, pg. 7 References Provided to Applicant None.

Answer: B

40. 000022G2.1.7 1 Current conditions:

• Unit 1 is operating at 100% power

• The following alarms on annunciator panel 1B are lit:

• A6 - CTMT PART PRESS -0.1 CH 1

• B6 - CTMT PART PRESS -0.1 CH 2

• Bulk Containment temperature on the Plant Computer shows a slowly rising trend.

• Three control rod drive mechanism (CRDM) fans are running.

• Three Containment Air Recirculation Fans (CARF) are running.

Based on the above conditions, which ONE of the following actions will clear the containment partial pressure alarms?

A. Place the in-service Containment Vacuum Pump control switch to OFF.

B. Increase Component Cooling (CC) water flow in Header A.

C. Increase cooling through the operating CC Chilled Water chiller.

D. Start an additional CRDM cooling fan.

K/A Containment Cooling Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior and instrumentation interpretation.

(CFR: 41.5/43.5/45.12/45.13) (RO - 4.4)

K/A Match Analysis The RO applicant is required to know the relation between containment partial pressure, containment temperature and the operating characteristics of the containment ventilation cooling and containment vacuum systems to determine the correct action.

Answer Choice Analysis A. INCORRECT. Plausible because the operating containment vacuum pump can cause partial pressure to decrease and it is an action directed by ARPs 1B-A6, -B6.

However, it would not explain the increasing trend on Containment temperature and the vacuum pump automatically stops when partial pressure is -0.1 from setpoint.

B. INCORRECT. Plausible because CC provides cooling to the CRDM coolers and can provide cooling to the CARFs. However, under these conditions (i.e., full power) the CC Chilled Water system would be aligned to provide cooling to the CARFs and CC header B provides cooling to Containment Ventilation, not CC header A.

C. CORRECT. Partial pressure can decrease as a result of either decreasing containment pressure or as a result of increasing containment temperature (i.e.

increasing Psat). So a decreasing partial pressure coupled with increasing containment temperature would mean more containment cooling was needed.

Increasing coolant flow through the operating CC Chilled Water chiller will cool the containment air passing through the CARFs and in turn cause the containment temperature to decrease and bring the partial pressure back to its setpoint. As containment temperature increases vapor pressure increases non-linearly. Partial pressure is determined by subtracting the actual containment pressure from the calculated vapor pressure to establish the partial pressure for the water vapor. The partial air pressure is determined by subtracting the partial pressure due to water vapor from actual containment pressure.

D. INCORRECT. Plausible because there are six CRDM coolers. However, there are three control switches with two CRDM fans controlled by each switch. Only one fan can be operated at a time using these switches. With three CRDM coolers operating it would not be possible to start an additional CRDM cooler.

Supporting References

1. Surry lesson plan ND-88.4-LP-6, "Containment Ventilation," rev. 9, Obj A & B, pp.

3-5.

2. Surry lesson plan ND-88.4-LP-5, "Containment Vacuum," rev. 12, Obj A, pp. 3-7.

3. 1B-A6, CTMT PART PRESS -0.1 CH 1, Rev. 7 References Provided to Applicant None.

Answer: C

41. 000026K2.01 1 Unit 1 conditions:

- Time = 1000: Large Break LOCA

- Time = 1002: Loss of Offsite Power Based on the above conditions, as the Emergency Buses energize, which one of the following states (1) which recirc spray pumps start first and (2) the reason for this action?

A. (1) Inside Recirc Pumps (2) Their spray header is full of water which results in spraying the containment sooner B. (1) Inside Recirc Pumps (2) Their starting current is larger so starting them earlier in the sequence will minimize the chance of overloading the DGs C. (1) Outside Recirc Pumps (2) Their spray header is full of water which results in spraying the containment sooner D. (1) Outside Recirc Pumps (2) Their starting current is larger so starting them earlier in the sequence will minimize the chance of overloading the DGs K/A Containment Spray: Knowledge of bus power supplies to the following: Containment spray pumps.

K/A Match Analysis Requires knowledge of how power is sequenced to the Containment Spray Pumps after power was lost.

Answer Choice Analysis A. INCORRECT. The Outside Containment Recirc Pumps are started first. 1st part is plausible because the Inside Containment Recirc Pumps have shorter distance from the pumps to the spray header. 2nd part is correct.

B. INCORRECT. The Outside Containment Recirc Pumps are started first. 1st part is plausible because the Inside Containment Recirc Pumps have shorter distance from the

pumps to the spray header. 2nd part is incorrect but plausible because assuming that this Pump will pump more water when starting, it will have a higher starting current.

C. CORRECT. The Outside Containment Recirc Pumps are started first because the headers are kept full of water by a check valve. This results in water spraying the containment as soon as the pump starts which limits containment pressure.

D. INCORRECT. 1st part is correct. 2nd part is incorrect but plausible because assuming that this Pump will pump more water when starting, it will have a higher starting current.

Supporting References ND-90.3-LP-7 Obj: D References Provided to Applicant None.

Answer: C

42. 000039A1.03 1 Unit 1 Initial Conditions (1000 hrs):

• The Reactor is at (add a percent power which coincides with where they would warm main steam lines)

• RCS temperature is 547 F

• Main steam lines are being heated using the MSIV bypass valves Current Conditions (1015 hours0.0117 days <br />0.282 hours <br />0.00168 weeks <br />3.862075e-4 months <br />):

• RCS temperature is 520 F Based on the current conditions, which one of the following states (1) whether a reactor trip is required, and (2) whether Technical Specification cooldown rate limits have been exceeded?

A. (1) Trip required.

(2) Cooldown rate limits exceeded.

B. (1) Trip required.

(2) Cooldown rate limits NOT exceeded.

C. (1) Trip NOT required.

(2) Cooldown rate limits exceeded.

D. (1) Trip NOT required.

(2) Cooldown rate limits NOT exceeded.

K/A Main and Reheat Steam A1.03 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MRSS controls including: Primary system temperature indications, and required values, during main steam system warm-up.

K/A Match Analysis The question requires knowledge of procedure actions (in this case the procedure is Tech Specs) as a result of exceeding the minimum temperature for criticality during the warming of the main steam lines. Therefore, the K/A is met because the applicant must monitor the change in RCS temp and know what actions to take based on that observed temperature change.

Answer Choice Analysis A. INCORRECT. The first part is correct. The second part is incorrect but plausible because the cooldown rate, if applied over 15 minutes, would be exceeded B. CORRECT. The minimum temp for criticality is 522 F. The cooldown causes RCS temp to be below this, which requires a reactor trip. The tech spec cooldown limit calculated over 15 minutes is in excess of the 100 F/hr Tech Spec limit; however, this would be a misapplication of the Tech Spec cooldown limits which are to be applied over a one hour period and not a 15 minute period. The second part is incorrect, but plausible because the applicant could misapply the cooldown rate limits over 15 minutes.

C. INCORRECT. The first part is not correct as discussed above, but it is plausible because not tripping, but simply closing MSIV bypass valves would be a method to lower the cooldown rate.

D. INCORRECT. The first part is not correct but plausible because the applicant may recognize that cooldown rates are not exceeded. The second part is correct.

Supporting References

• Tech Spec 3.1.E, Reactor Coolant System - Minimum Temperature for Criticality.

• Tech Spec 3.1.B, Reactor Coolant System - Heatup and Cooldown.

References Provided to Applicant

None.

The licensee will need to help supply supporting material to justify whether a reactor trip is required. Verbally, the licensee stated that a reactor trip was required, but the supporting references tend to state that being below the limit is not allowed, but then does not state what actions are required because of being below the limit. Either B or D is correct, pending research by the licensee.

Answer: B

43. 000059A2.11 1 Unit 1 Initial Conditions:

• Power = 15% and stable for a chemistry hold.

• PT-MS-446 is the selected turbine first stage (impulse) pressure channel.

• All Feed Regulating Valves (FRV) are controlling steam generator (S/G) levels in AUTO.

Current conditions:

• Time = 1440. PT-MS-446 fails high over a two minute period.

• Time = 1442. The operator realizes that PT-MS-446 has failed and begins to take the appropriate actions.

Based on the current conditions, which one of the following (1) is the response of the steam generator level control system to the PT-MS-446 failure IF no operator actions are taken, AND (2) the required operator actions to restore the steam generator level control system, in accordance with 0-AP-53.00, LOSS OF VITAL INSTRUMENTATION/CONTROLS?

A. (1) S/G levels will stabilize at 44%.

(2) VERIFY PT-MS-447 indications are NORMAL, then Select PT-MS-447. Once S/G levels are stable, Place all FRVs to MANUAL and then back to AUTO to remove windup from the controllers.

B. (1) S/G levels will stabilize at 44%.

(2) Place all FRVs to MANUAL, then Select PT-MS-447, then Place the FRVs back to AUTO.

C. (1) S/G levels will increase until the High S/G level trip setpoint is reached.

(2) Place all FRVs to MANUAL, then Select PT-MS-447, then Place the FRVs back to AUTO.

D. (1) S/G levels will increase until the High S/G level trip setpoint is reached.

(2) VERIFY PT-MS-447 indications are NORMAL, then Select PT-MS-447. Once S/G levels are stable, Place all FRVs to MANUAL and then back to AUTO to remove windup from the controllers.

K/A 059 Main Feedwater 059A2.11 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: Failure of feedwater control system.

(CFR 41.5 / 43.5 / 45.3 / 45.13) (RO - 3.0)

K/A Match Analysis This question matches the K/A statement by requiring the RO applicant to correctly predict the impact of a failure of the feedwater control system (a failure of the selected impulse pressure channel, which is an input into SGWLC control), and to the demonstrate knowledge of the correct actions to take in accordance with plant procedures to mitigate the effects of the feedwater control system failure.

Answer Choice Analysis A. INCORRECT. Part (1) of this distractor is correct, S/G levels will stabilize at 44%,

which is the SGWLC reference level setpoint from 20-100% power. The Surry lesson plan on SGWLC states the following for the selected Pimp channel failing high: If turbine power was <20% and the FRV was in automatic, the FRV will throttle open to increase S/G level to 44%. Part (2) of this distractor is incorrect. The same lesson plan states: The required operator action for a Pimp failure either high or low will be to place the FRVs in manual and then select the unaffected Pimp channel for SGWLC.

The operator can then place the FRVs back in automatic. This statement is a summary of actions contained in Surry procedure 0-AP-53.00, specifically in Attachment 2 of this procedure. The part (2) distractor is plausible, because taking these actions will restore S/G levels, and will also remove windup from the controllers; however, the sequence is incorrect as laid out in the AP procedure.

B. CORRECT. See analysis of A above.

C. INCORRECT. See analysis of A above. Part (1) is incorrect, but plausible, because a failure of the pressure channel high does cause S/G levels to go upit just clips the signal at the high level setpoint (44%), instead of increasing without bound.

D. INCORRECT. See analysis of A and C above. Both parts (1) and (2) incorrect.

Supporting References

1. Surry Lesson Plan ND-93.3-LP-8, SG WATER LEVEL CONTROL SYSTEM, rev. 7

dtd 09/27/04.

2. Surry Procedure 0-AP-53.00, LOSS OF VITAL INSTRUMENTATION/CONTROLS, rev. 15.

References Provided to Applicant None.

Answer: B

44. 000059A4.03 1 Unit 1 Initial Conditions:

• Power = 96%.

• The unit has operated at full power for 356 days of continuous operation, and has been in a power coastdown for the past four days.

• RCS Boron concentration is 20 ppm.

Current conditions:

• Time = 0004. The operations team has just commenced a power reduction in accordance with 1-GOP-2.1, UNIT SHUTDOWN, POWER DECREASE FROM ALLOWABLE POWER TO LESS THAN 30% REACTOR POWER, in order to begin a refueling outage.

• An operator inadvertently opens 1-FW-FCV-150A, 1A Main Feed Pump Recirc Valve.

Based on the current conditions, which one of the following completes the below statement, in accordance with 1-GOP-2.1?

When the Main Feed Pump Recirc Valve opens, Feedwater temperature will (1)

______________ that will add (2) _______________________ to the core as compared to the exact same situation at beginning of life (BOL).

A. (1) decrease (2) a larger amount of positive reactivity B. (1) decrease (2) a smaller amount of positive reactivity C. (1) increase (2) a larger amount of negative reactivity

D. (1) increase (2) a smaller amount of negative reactivity K/A 059 Main Feedwater 059A4.03 Ability to manually operate and monitor in the control room: Feedwater control during a power increase and decrease.

(CFR 41.7 / 45.5 to 45.8) (RO - 2.9)

K/A Match Analysis This question matches the K/A statement by requiring the RO applicant to correctly apply knowledge of the integrated plant response to an inadvertent operation of the main feed recirc valve during a downpower situation.

Answer Choice Analysis A. CORRECT. A NOTE before step 5.3.8 of 1-GOP-2.1 states the following: When a Main Feed Pump Recirc Valve opens, a decrease in FW temperature will occur that can add positive reactivity to the core. The magnitude of the reactivity change is dependent on the time in core life and the value of the Moderator Temperature Coefficient.

Therefore, FW temperature will decrease and part (1) is correct. Because the given situation is at the extreme EOL, the magnitude of the MTC is at its greatest point (most negative) as compared to BOL, when its magnitude is smaller. Therefore, the reactivity effect is greater than the effect of the same transient at BOL, and part (2) is also correct.

B. INCORRECT. See analysis of A above. Part (1) is correct, part (2) incorrect. A smaller magnitude of reactivity change is plausible if the applicant wrongly believes that the magnitude of the MTC is less at EOL than at BOL.

C. INCORRECT. See analysis of A above. Part (1) is incorrect, part (2) would be correct if part (1) was correct.

D. INCORRECT. See analysis of A and B above. Both parts (1) and (2) incorrect.

Supporting References

1. Surry Procedure 1-GOP-2.1, UNIT SHUTDOWN, POWER DECREASE FROM ALLOWABLE POWER TO LESS THAN 30% REACTOR POWER, rev. 34. Especially NOTE before step 5.3.8.

2. Surry Procedure 1-DRP-003, CURVE BOOK, rev. 90. Attachments 26 and 46.

References Provided to Applicant None.

Answer: A

45. 000061A3.03 1 Initial Conditions:

• Reactor trip occurred.

• AFW system received an auto-start signal on low S/G levels.

• 1FW-43-1-AFW-E, H Train Auto Open Enable Selector Switch, placed in the Disable Selected position.

• 1-FW-43-3-AFW-S, H Train Disable Selector Switch placed in the A Steam Generator position. (Need more complete switch description - lesson plan does not describe the labeling on the switch.)

• ALL of the steam generator level control valves (1-MOV-FW-151A thru F) were manually closed.

• No reset actions were taken.

Current conditions:

• Steam generator NR levels are 63%.

• An automatic SI actuation has been received.

Which one of the following choices completes the statement below concerning response of the MOV-FW-151 valves to the SI actuation?

(1) remains closed and all other MOV-FW-151 valves will (2) in response to the SI actuation.

A. (1) 1-MOV-FW-151E (2) automatically open B. (1) 1-MOV-FW-151F (2) remain closed C. (1) 1-MOV-FW-151F (2) automatically open D. (1) 1-MOV-FW-151E (2) remain closed K/A

Auxiliary Feedwater System Ability to monitor automatic operation of the AFW, including: AFW Steam Generator level control on automatic start.

(CFR: 41.7/45.5) (RO - 3.9)

K/A Match Analysis The RO applicant is required to determine whether the steam generator level control valves (1-FW-MOV-151 valves) will automatically open when a subsequent AFW auto-start signal (SI initiation) is received and which level control valve for A S/G will be disabled by the H Train Disable Selector Switch.

Answer Choice Analysis A. INCORRECT. Plausible because 1-MOV-FW-151E is the correct level control valve for A S/G that is disabled. The second half of the response is plausible because the remaining level control valves would open if the original AFW auto-start signal had been reset.

B. INCORRECT. Plausible because1-MOV-FW-151F is the level control valve for A S/G that would be disabled if J Train Disable Selector Switch were placed in the A S/G position. The second half of the response is correct.

C. INCORRECT. Plausible because1-MOV-FW-151F is the level control valve for A S/G that would be disabled if J Train Disable Selector Switch were placed in the A S/G position. The second half of the response is plausible because the remaining level control valves would open if the original AFW auto-start signal had been reset.

D. CORRECT. The H Train Disable Selector Switch is used to individually select the level control valve for each S/G on the A AFW header. For A S/G this would disable 1-MOV-FW-151E. Also, the level control valves will not receive an open signal on a subsequent auto-start signal (SI actuation) without resetting all signals that feed the 45-second open signal to the level control valves.

Supporting References

1. Surry lesson plan ND-89.3-LP-4, "Auxiliary Feed System," rev. 26, Obj F, pgs.

15-17.

References Provided to Applicant None.

Answer: D

46. 000061K5.05 1

Current conditions on Unit 1:

• 100% power

• B AFW header temperature indicates 168°F and increasing on 1-FW-TI-160B.

Which ONE of the following describes:

(1) the actions that should be taken in accordance with 1-AP-21.01, Response to AFW Check Valve Backleakage AND (2) the reason the action is taken?

A. (1) Close MOV-FW-151A, -151C, and -151E.

(2) To prevent S/G blowdown in the event of a piping rupture upstream of the valve.

B. (1) Close MOV-FW-151B, -151D, and -151F.

(2) To prevent S/G blowdown in the event of a piping rupture upstream of the valve.

C. (1) Close MOV-FW-151A, -151C, and -151E (2) To prevent vapor binding of the AFW headers and pump casings.

D. (1) Close MOV-FW-151B, -151D, and -151F (2) To prevent vapor binding of the AFW headers and pump casings.

Auxiliary Feedwater System Knowledge of the operational implications of the following concepts as they apply to the AFW: Feed line voiding and water hammer.

(CFR: 41.5/45.7) (RO - 2.7)

K/A Match Analysis The RO applicant is required to recognize that the elevated AFW header temperature indicates leakage around the stop check valves and the operational implications would be the potential for voiding of the AFW headers and pump casings.

Answer Choice Analysis A. INCORRECT. Plausible because the listed valves are the 151 MOVs for each of the steam generators on the A AFW header. Also, prevention of S/G blowdown is the purpose FW check valves on the feedwater headers.

B. INCORRECT. Plausible because these are the correct 151 MOVs that are closed to isolate the B AFW header from the steam generators. Also, prevention of S/G

blowdown is the purpose FW check valves on the feedwater headers.

C. INCORRECT. Plausible because the listed valves are the 151 MOVs for each of the steam generators on the A AFW header. These valves are closed because the elevated temperature indicates leakage around the stop check valves. This can lead to voids developing in the AFW headers and pump casings and potentially result in vapor binding of the AFW pumps.

D. CORRECT. When AFW Header temperature exceeds 165°F, Step 4 of 1-AP-21.01 directs the user to close either the individual 151 MOVs or to close all the 151 MOVs on the affected AFW header. The listed valves are the 151 MOVs for each of the steam generators on the B AFW header.

Supporting References

1. Surry lesson plan ND-89.3-LP-3, Main Feedwater System, rev. 21, p. 5

2. Surry lesson plan ND-89.3-LP-4, "Auxiliary Feed System," rev. 26, Obj C, p. 17.

3. 1-AP-21.01, Response to AFW Check Valve Backleakage, Rev. 5, pg. 2 References Provided to Applicant None.

Answer: D

47. 000062A1.01 1 Unit 1 conditions:

- Rx Trip with loss of offsite power occurred

- EDG 1 is powering Bus 1H

- EDG 3 is powering Bus 1J CH-P-1A is running CH-P-1C is aligned to Bus 1H AP-10.07, Loss of Unit 1 Power, Attachment 4, Emergency Bus Load Alignment, is in progress Based on the above conditions, which one of the following states (1) the action required to be taken with charging pump 1-CH-P-1B in accordance with 1-AP-10.07, Attachment 4, and (2) the consequence if this action is not taken?

A. (1) Start 1-CH-P-1B (2) Subsequent HI HI CLS actuation could overload EDG 3 B. (1) Start 1-CH-P-1B

(2) Subsequent HI HI CLS actuation could overload EDG 1 C. (1) Place 1-CH-P-1B in PTL (2) Subsequent HI HI CLS actuation could overload EDG 3 D. (1) Place 1-CH-P-1B in PTL (2) Subsequent HI HI CLS actuation could overload EDG 1 K/A Electrical Distribution: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls including: Significance of D/G load limits.

K/A Match Analysis Requires knowledge of Emergency Bus load limit when being powered from the EDGs.

Answer Choice Analysis A. CORRECT. Per 1-AP-10.07, If EDG 3 is the sole source of power to Emergency Bus 1J, start 1-CH-P-1B. The purpose of Attachement 4 is stated as: To limit EDG loading so that a subsequent HI HI CLS actuation willnot overload the EDG.

B. INCORRECT. 1st part is correct. 2nd part is incorrect because 1-CH-P-1B is not powered from EDG 1. Plausible because EDG 1 does does provide power 1-CH-P-1A and 1C.

C. INCORRECT. Per 1-AP-10.07, If EDG 3 is the sole source of power to Emergency Bus 1J, start 1-CH-P-1B. 1st part is plausible because if -CH-P-1C was lined up to the same bus and was operating, it would be correct. 2nd part is correct.

D. INCORRECT. Per 1-AP-10.07, If EDG 3 is the sole source of power to Emergency Bus 1J, start 1-CH-P-1B. 1st part is plausible because if -CH-P-1C was lined up to the same bus and was operating, it would be correct.2nd part is incorrect because 1-CH-P-1B is not powered from EDG 1. Plausible because EDG 1 does does provide power 1-CH-P-1A and 1C.

Supporting References 1AP-10.07 Attachment 4, Emergency Bus Load Alignment References Provided to Applicant None.

Answer: A

48. 000062K3.01 1 Unit 1 conditions:

- Unit 1 semi-vital bus is faulted

- Unit 1 tripped aproximately one hour ago during the down power required due to the faulted semi-vital bus Which one of the following actions are required to maintain Tave at 547oF during repair of the semi-vital bus?

A. Dump steam using the steam dumps.

B. Dump steam using local operation of the SG PORVs.

C. Dump steam using auto controller operation of the SG PORVs.

D. Dump steam using manual controller operation of the SG PORVs.

K/A 062K3.01 AC Electrical Distribution Knowledge of the effect that a loss or malfunction of the AC distribution system will have on the following: Major system loads.

K/A Match Analysis The question requires knowledge of how to maintain Tave in the presence of a loss of AC. The local dumping of steam is an effect of the AC malfunction.

Answer Choice Analysis A. Incorrect. Plausible and incorrect because IAW Page 3 of 1-AP-10.05, dumps will be available in Tave mode for only 30 minutes after the loss of power.

B. Correct. Local operation is directed by 1-AP-10.05. Att 4 provides direction.

C. Incorrect. Local operation is the only option after 30 minutes. Plausible because this is a viable option until the UPS runs out of power at about 30 minutes.

D. Incorrect. Local operation is directed by 1-AP-10.05. Att 4 provides direction.

C. Incorrect. Local operation is the only option after 30 minutes. Plausible because this is a viable option until the UPS runs out of power at about 30 minutes.

Supporting References 1-AP-10.05, Loss of Semi-Vital Bus, Rev. 25.

References Provided to Applicant None.

Question History Surry 2002 Exam Answer: B

49. 000063G2.2.44 1 Unit 1 Initial Conditions:

• Power = 100%.

Current conditions:

• Reactor trips with no operator action.

• Turbine trips.

• Generator output breakers open, but the generator exciter field breaker does NOT open automatically when the turbine initially trips.

• In the absence of operator action, the exciter field eventually automatically secures on Volts/Hertz protection.

• #1 EDG automatically starts but the #1 EDG output breaker does NOT operate either automatically or remotely from the control panels.

Based upon the current conditions, which one of the following describes (1) what event caused the above indications, AND (2) the impact on breaker overcurrent protection?

A. (1) Loss of A DC Bus.

(2) A DC Bus 480V breakers AND A DC Bus 4160V breakers will trip on overcurrent.

B. (1) Loss of A DC Bus.

(2) A DC Bus 480V breakers will trip on overcurrent. A DC Bus 4160V breakers will NOT trip on overcurrent.

C. (1) Loss of B DC Bus.

(2) B DC Bus 480V breakers AND B DC Bus 4160V breakers will trip on overcurrent.

D. (1) Loss of B DC Bus.

(2) B DC Bus 480V breakers will trip on overcurrent. B DC Bus 4160V breakers will NOT trip on overcurrent.

K/A

063 DC Electrical Distribution 063G2.2.44 Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

(CFR 41.5 / 43.5 / 45.12) (RO - 4.2)

K/A Match Analysis This question matches the K/A statement by requiring the RO applicant to correctly interpret the control room indications to determine whether an operationally valid scenario is a loss of A DC Bus or a loss of B DC Bus, and then demonstrate knowledge of how the casualty impacts overall emeregency breaker operations.

Answer Choice Analysis A. INCORRECT. Part (1) of this distractor is correct. The indications that are given are exactly as written in Surry lesson plan ND-90.3-LP-6, 125 VDC DISTRIBUTION, for Expected plant response of a loss of A DC bus while at power (pages ND-90.3-AIA-6.2 page 1 and 2). Part (2) of this distractor is incorrect. The NOTE on page 2 of 9 in the lesson plan states: the 4160V breakers do not have overcurrent protection. The 480V breakers overcurrent protection is mechanical (excessive current through an overcurrent coil pulls up on a tripper bar inside the breaker) so they will trip.

B. CORRECT. See analysis of A above.

C. INCORRECT. See analysis of A above.

D. INCORRECT. See analysis of A above.

Supporting References

1. Surry lesson plan ND-90.3-LP-6, 125VDC DISTRIBUTION, rev. 16. Especially section for for Expected plant response of a loss of A DC bus while at power (pages ND-90.3-AIA-6.2 page 1 and 2).

References Provided to Applicant None.

Answer: B

50. 000063K1.02 1 Unit 1 Initial Conditions:

• Refueling outage is in progress.

• Reactor is completely de-fueled.

• Electrical systems from the 480V Emergency Buses to the Vital Buses are in a normal at-power line-up.

Current conditions:

• Time = 0215. Battery 1B is disconnected from DC Bus 1B for maintenance.

• Time = 0220. An electrical fault causes UPS 1B1 to de-energize.

Based on the current conditions, with no operator actions taken, which one of the following is the status of Vital Buses as of Time = 0221 ?

Vital Buses Vital Buses 1-II and 1-IIA 1-IV and 1-IVA A. De-energized De-energized B. Energized De-energized C. Energized Energized D. De-energized Energized K/A 063 DC Electrical Distribution 063K1.02 Knowledge of the physical connections and/or cause-effect relationships between the DC electrical system and the following systems: AC electrical system.

(CFR 41.2 to 41.9 / 45.7 to 45.8) (RO - 2.7)

K/A Match Analysis This question matches the K/A statement by requiring the RO applicant to recall in a straight-forward fashion the one-line diagram from the VITAL AND EMERGENCY DISTRIBUTION, and to correctly determine which vital AC buses would lose power based on operationally valid problems with the DC electrical system.

Answer Choice Analysis A. INCORRECT. Based on the provided diagram ND-90.3-H/T-5.2, VITAL AND EMERGENCY DISTRIBUTION, when Battery 1B, DC Bus 1B, and UPS 1B1 lose power, Vital Buses 1-II and 1-IIA will lose power. Vital Bus 1-III and 1-IIIA will remain energized from UPS 1A2, and Vital Bus 1-IV and 1-IVA will remain energized by UPS

1B2 via MCC 1J1-1 or MCC 1J1-2. All other choices are plausible combinations of these buses.

B. INCORRECT. See analysis of A above.

C. INCORRECT. See analysis of A above.

D. CORRECT. See analysis of A above.

Supporting References

1. Surry Lesson Plan ND-90.3-LP-5, VITAL AND SEMI-VITAL BUS DISTRIBUTION, rev. 15. Especially the diagram that is provided as handout 5.2, VITAL AND EMERGENCY DISTRIBUTION.

References Provided to Applicant None.

Answer: D

51. 000064K6.08 1 Current conditions:

• Unit 1 has experienced a loss of offsite power.

• #1 EDG has started but failed to load.

• A field operator reports the following:

• The exciter field circuit breaker on #1 EDG is tripped.

Based on the current conditions, which one of the following describes:

(5) the impact this will have on the ability to transfer fuel oil to the #1 EDG and (6) the minimum volume necessary to meet the Technical Specification Day Tank requirements?

The fuel oil system will no longer be able to transfer fuel to the (1) .

The Day Tank volume must be greater than or equal to (2) .

A. (1) Base Tank, but will continue to transfer fuel to the Wall Tank.

(2) 290 gallons.

B. (1) Base Tank, but will continue to transfer fuel to the Wall Tank.

(2) 375 gallons.

C. (1) Base Tank and it will be unable to transfer fuel to the Wall Tank.

(2) 375 gallons.

D. (1) Base Tank and it will be unable to transfer fuel to the Wall Tank.

(2) 290 gallons.

K/A Emergency Diesel Generator Knowledge of the effect of a loss or malfunction of the following will have on the ED/G system: Fuel oil storage tanks (CFR: 41.7/45.7) (RO - 3.2)

K/A Match Analysis The RO applicant is required to recognize that a major portion of the Fuel Storage volume will be lost and that only the volume of the Base Tank will be available to the EDG. In addition, the RO applicant is required to know that the Technical Specification minimum Day Tank volume is based on the combined volumes of the Wall and Base Tanks.

Answer Choice Analysis A. INCORRECT. Plausible because the pumps that transfer fuel from the Wall Tank to the Base Tank are powered by the EDG field. So with exciter field breaker open power would not be available to these pumps. However, without an exciter field the EDG will not load. The fuel transfer pumps from the underground storage tank are fed from the emergency bus (H-1) supplied by the EDG. The system would be unable to replenish either the Wall Tank or the Base Tank. Also, the T.S. Day Tank volume requirement is 290 gallons.

B. INCORRECT. Plausible because the pumps that transfer fuel from the Wall Tank to the Base Tank are powered by the EDG field. So with exciter field breaker open power would not be available to these pumps. However, without an exciter field the EDG will not load. The fuel transfer pumps from the underground storage tank are fed from the emergency bus (H-1) supplied by the EDG. The system would be unable to replenish either the Wall Tank or the Base Tank. The second half of the question is plausible because that is the low level auto-start signal for the fuel transfer pumps to the Base Tank.

C. INCORRECT. The first half of the answer is correct. The second half of the question is plausible because that is the low level auto-start signal for the fuel transfer pumps to the Base Tank.

D. CORRECT. See the discussion for choice A above.

Supporting References

1. Surry lesson plan ND-90.3-LP-1, "Emergency Diesel Generator," rev. 20, Objectives B & M , pgs. 9, 47.

2. Technical Specifications, 3.16 References Provided to Applicant None.

Answer: D

52. 000073G2.2.22 1 Unit 1 plant conditions:

Refueling in progress One Ventilation vent stack particulate and gas monitor (1-VG-RM-131-1) loses power Based on the above conditions, which one of the following states the actions required by Technical Specifications / Technical Requirements Manual?

A. Fuel handling is allowed to continue with one operable vent stack particulate and gas monitor.

B. Fuel handling is allowed to continue if airborne samples are performed every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> while fuel handling is in progress.

C. Fuel handling is allowed to continue for up to one hour while repair attempts are made on the detector.

D. Fuel handling must cease immediately and no operations which increase the reactivity of the core shall be made.

K/A Process Radiation monitoring: G2.2.22 Process Radiation Monitoring System:

Knowledge of limiting conditions for operations and safety limits.

K/A Match Analysis Requires applicant to know how various Rad Monitors will indicate when exceeding plant limits.

Answer Choice Analysis A. Incorrect: Plausible because various instrumentation Tech Specs allow one instrument to me inoperable and still allow operation.

B. Incorrect: Plausible because waste gas holdup discharge is allowed with this method when a monitor is out of service.

C. Incorrect: Plausible because one detector is still operable and it allows the operator time to place fuel in a safe location.

D. Correct: Per TR 3.3.3: A.1 As specified in TS 3.10.C, REFUELING OPERATIONS or irradiated fuel movement in the Fuel Building shall cease, work shall be initiated to correct the conditions so that TS 3.10.A.3 or TS 3.10.B.1 are met, and no operations which increase the reactivity of the core shall be made.

Supporting References TS 3.10.B.1 Irradiated Fuel Movement TRM 3.3.3 Radiation Monitor Requirements References Provided to Applicant none Answer: D

53. 000076A3.02 1 Current conditions:

• Unit 1 has experienced a large break LOCA.

• Containment pressure peaked at 24.5 psia.

Based on the above current conditions, which one of the following components will have their supporting Service Water equipment repositioned from the equipments original position before the LOCA?

A. Recirculation Spray heat exchangers.

B. Component Cooling Water heat exchangers.

C. Bearing Cooling Water heat exchangers.

D. Charging Pump oil coolers.

K/A

Service Water Ability to monitor automatic operation of the SWS, including: Emergency heat loads.

(CFR: 41.7/45.5) (RO - 3.7)

K/A Match Analysis The RO applicant is required to recognize the systems that will automatically realign as a result of a Hi-Hi Containment condition.

Answer Choice Analysis A. CORRECT. A Hi-Hi CLS signal will cause 1-MOV-103, -104 and -105 valves on each Recirculation Spray heat exchangers to open to allow Service Water flow to the heat exchangers.

B. INCORRECT. Plausible because the inlet and outlet SW valves on the CC heat exchangers close on Hi-Hi CLS signal, but only in coincidence with a blackout (LOOP) signal.

C. INCORRECT. Plausible because the supply SW valves on the Bearing Cooling heat exchangers close on Hi-Hi CLS signal, but only in coincidence with a blackout (LOOP) signal.

D. INCORRECT. Plausible because the Charging Pump Service Water System does receive an automatic signal, but it results from a low discharge pressure signal, which will autostart the standby pump.

Supporting References

1. Surry lesson plan ND-89.5-LP-2, "Service Water System," rev. 27, Obj F, pp. 15, 18-19.

2References Provided to Applicant None.

Answer: A

54. 000078K4.02 1 Unit 1 initial plant conditions:

Reactor is at Hot Shutdown conditions The Containment Instrument Air System will be required to be supplied from its back up air source Based on the above conditions, which one of the following states (1) the air system that

acts as a backup to the Containment Instrument Air System and (2) if the operator stationed at the containment isolation valve is required to be in constant communication with the control room per TS 3.8 Containment?

A. (1) Service Air System (2) No B. (1) Service Air System (2) Yes C. (1) Instrument Air System (2) No D. (1) Instrument Air System (2) Yes K/A Instrument Air. Knowledge of IAS design feature(s) and/or interlock(s) which provide for the following: Cross-over to the other air systems.

K/A Match Analysis Requires applicant to know how the Instrument Air system connects to the Containment Air System.

Answer Choice Analysis A. Incorrect: Instrument Air is the backup to the Containment Instrument Air system.

Plausible because the Service Air system backs up the Instrument Air system. 2nd part is plausible because TS 3.8 also requires the operator to be at the containment isolation valve which the applicant may think is sufficient.

B. Incorrect: Instrument Air is the backup to the Containment Instrument Air system.

Plausible because the Service Air system backs up the Instrument Air system. 2nd part is correct per 3.8 Containment.

C. Incorrect: 1st part is correct. Instrument Air is the backup to the Containment Instrument Air system. 2nd part is plausible because TS 3.8 also requires the operator to be at the containment isolation valve which the applicant may think is sufficient.

D. Correct: Instrument Air is the backup to the Containment Instrument Air system.

2nd part is correct per 3.8 Containment: The opening of manual or deactivated automatic containment isolation valves on an intermittent basis under administrative control includes the following considerations: (1) stationing an operator, who is in constant communication with the control room, at the valve controls.

Supporting References ND-92.1-LP-1 Obj E

TS 3.8 References Provided to Applicant none Answer: D

55. 000103K3.01 1 Unit 1 initial conditions:

Reactor heat up is in progress RCS Temp = 245 oF The Containment Ventilation Purge System outside exhaust containment isolation valve is reported as open.

Based on the above conditions, which one of the following states: (1) TS required actions for the containment isolation valve and (2) whether heatup to greater than 547 oF is allowed by Tech Specs after the required actions for the containment isolation valve has been completed?

A. (1) Ensure valve is locked and sealed within one hour (2) Yes B. (1) Ensure valve is locked and sealed within one hour (2) No C. (1) The valve is allowed to be open if administrative controls are established (2) Yes D. (1) The valve is allowed to be open if administrative controls are established (2) No K/A Containment System: Knowledge of the effect that a loss of malfunction of the containment system will have on the following: Loss of containment integrity under shutdown conditions.

K/A Match Analysis Requires applicant to know how a containment isolation valve affects containment integrity and what actions are required.

Answer Choice Analysis

A. Correct: 1st part is correct (see B). 2nd part is correct.

B. Incorrect: TS 3.8.A.2 The inside and outside isolation valves in the Containment Ventilation Purge System shall be locked, sealed, or otherwise secured closed whenever the Reactor Coolant System temperature exceeds 200 oF. Heating up to 547 oF constitutes a mode change to Hot Shutdown which is allowed if the actions are completed.

C. Incorrect: 1st part is incorrect but plausible because intermittent opening of containment isolation valves is allowed but not on Containment Purge valves.

D. Incorrect: 1st part is incorrect but plausible because intermittent opening of containment isolation valves is allowed but not on Containment Purge valves.

Supporting References TS 1.0-1 Definitions TS 3.8 Containment References Provided to Applicant none Licensee to verify that a mode change is not allowed while in an LCO.

Answer: A

56. 002K1.01 1 Unit 1 Current Conditions:

• Reactor Vessel Disassembly is in progress, and the Refueling team is ready to flood up the Reactor Cavity.

• Testing of Safety Injection check valves is NOT required.

Based on the current conditions, which one of the following completes the below statements in accordance with 1-OP-FH-001, CONTROLLING PROCEDURE FOR REFUELING?

(1) The preferred method for cavity fill is from the Refueling Water Storage Tank (RWST) through the Reactor Coolant System ____________________ .

(2) When the cavity is filled to between 26 and 27 feet, the final minimum required RWST level is ____________________ .

A. (1) cold legs (2) 22%

B. (1) cold legs (2) 12%

C. (1) hot legs (2) 22%

D. (1) hot legs (2) 12%

K/A 002 Reactor Coolant 002K1.01 Knowledge of the physical connections and/or cause-effect relationships between the RCS and the following systems: RWST.

(CFR 41.2 to 41.9) (RO - 3.7)

K/A Match Analysis This question matches the K/A statement by requiring the RO applicant to remember in a straight-forward fashion that it is preferred to fill the cavity via the hot legs to prevent introduction of crud into the cavity; also to remember that the RWST is checked at 85%

level to ensure a minimum 22% level when the cavity is full.

Answer Choice Analysis A. INCORRECT. Part (1) of the distractor is incorrect. The Surry lesson plan for Unit Refueling Overview states the following: Cavity fill is normally done through the safety injection hot legs (high head or low head). Fill using the cold legs is permitted only to test the check valves. If the cavity is filled using the cold legs crud is flushed from the reactor vessel reducing water clarity and raising radiation levels. Part (1), using cold legs, is plausible if the applicant mis-remembers the reasons for the preferred flow path, or if the applicant wrongly missed the bullet in the question stem that stated check valve testing was NOT to be performed. Part (2) of this distractor is correct. The RWST level is filled to at least 85% before cavity fill, in order to ensure at least 22% RWST level when the cavity fill operation is secured.

B. INCORRECT. Parts (1) and (2) incorrect. See analysis of A above. The answer for part (2) is 12% RWST level, which is a plausible wrong misconception given that the correct part (2) answer is 22%.

C. CORRECT. See analysis of A above.

D. INCORRECT. Part (1) correct, part (2) incorrect. See analysis of A and 'B' above.

Supporting References

1. Surry Lesson Plan ND-92.5-LP-1, UNIT REFUELING OVERVIEW, rev. 18.

2. Surry Procedure 1-OP-FH-001, CONTROLLING PROCEDURE FOR REFUELING, rev. 23.

References Provided to Applicant None.

Answer: C

57. 011A4.04 1 Initial conditions:

• Unit 1 is at 100% power.

• The PRZR LVL - CH SEL Switch is in the CH 3 & 2 position.

• Pressurizer backup heater bank B is energized.

Current conditions:

• Pressurizer level transmitter 1-RC-LT-1461 fails upscale.

Based on current conditions, which one of the following statements describes the effect on pressurizer level control and on the pressurizer heaters?

The pressurizer level control system A. will continue to maintain pressurizer level at program level and the remaining backup heater banks will energize.

B. is required in accordance with 0-AP-53.00, Loss of Vital Instrumentation/Controls, to be placed in manual to adjust pressurizer level back to the program level setting and the remaining backup heater banks will energize.

C. will continue to maintain pressurizer level at program level, however pressurizer backup heater bank B will de-energize.

D. is required in accordance with 0-AP-53.00, Loss of Vital Instrumentation/Controls, to be placed in manual to adjust pressurizer level back to the program level setting, however pressurizer backup heater bank B will de-energize.

K/A Pressurizer Level Control

Ability to manually operate and/or monitor in the control room: Transfer of PZR LCS from automatic to manual.

(CFR: 41.7/45.5 to 45.8) (RO - 3.2)

K/A Match Analysis The RO applicant is required to apply the knowledge that the selected level transmitter (CH III) to the upper control channel will provide the input to the pressurizer level control system and to the pressurizer heaters to energize on 5% level deviation signal. An upscale failure of Channel III pressurizer level transmitter, which corresponds to 1-RC-LT-1461, will require operator action to place the pressurizer level control system in manual to control pressurizer level and will result in energizing the pressurizer heaters.

Answer Choice Analysis A. INCORRECT. Plausible because the pressurizer level control system would continue to automatically control pressurizer level if the failed transmitter were assigned to the lower control channel. In addition, the lower control channel signal also provides a signal to the pressurizer heaters, however, it will cause the heaters to de-energize on low pressurizer level rather than energize on a deviation of pressurizer level above program level.

B. CORRECT. Since Channel 3 (1-RC-LT-1461) is selected to the upper control channel then an upscale failure of this transmitter will cause the pressurizer level controller to reduce the demand signal to the charging system. The pressurizer level controller will need to be placed in manual to control pressurizer level. Also, the level transmitter selected to the upper control channel provides the signal to the pressurizer heaters to energize when pressurizer level increases 5% above program level.

C. INCORRECT. Plausible because the pressurizer level control system would continue to automatically control pressurizer level if the failed transmitter were assigned to the lower control channel. In addition, the lower control channel does not provide a signal to the pressurizer heaters on a 5% level deviation above program level.

D. INCORRECT. Plausible because the first half of the response is correct. The second half is plausible if a misconception exists as to which control channel provides the signal to energize the pressurizer heaters.

Supporting References

1. Surry lesson plan ND-93.3-LP-7, "Pressurizer Level Control System," rev. 10, Objectives B and C, pp. 5, 9.

2. 0-AP-53, Loss of Vital Instrumentation/Controls, rev. 15, pg. 10.

References Provided to Applicant None.

Answer: B

58. 014K5.02 1 Unit 2 Initial Conditions (0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br />):

- Startup and power escalation are in progress.

- Reactor power is 25%.

- All control rods bank demand positions have remained aligned within 6 steps of their IRPI indication.

Unit 2 Current Conditions (0830 hours0.00961 days <br />0.231 hours <br />0.00137 weeks <br />3.15815e-4 months <br />):

- Reactor power is 35% when control rod F8 slips inward and its IRPI indicates 14 steps lower than its bank demand position Based on current conditions, which one of the following describes (1) whether the conditions of Tech Spec LCO 3.12.E, ROD POSITION INDICATION SYSTEM AND BANK DEMAND POSITION INDICATION SYSTEM, are currently met, and (2) required actions, if any, in accordance with 0-AP-1.01, CONTROL ROD MISALIGNMENT?

A. (1) Conditions of LCO 3.12.E ARE currently met.

(2) Shutdown margin requirements ARE required to be verified within one hour.

B. (1) Conditions of LCO 3.12.E ARE currently met.

(2) Shutdown margin requirements ARE NOT required to be verified within one hour.

C. (1) Conditions of LCO 3.12.E ARE NOT currently met.

(2) Shutdown margin requirements ARE required to be verified within one hour.

D. (1) Conditions of LCO 3.12.E ARE NOT currently met.

(2) Shutdown margin requirements ARE NOT required to be verified within one hour.

K/A: 014K5.02 Rod Position Indication Knowledge of the operational implications of the following concepts as they apply to RPIS:

RPIS independent of demand position.

KA MATCH ANALYSIS:

Knowledge of the operational implications of the misaligned rod is required to arrive at the correct answer. The applicant must know that the rod can be misaligned for up to one hour before the LCO is not met.

ANSWER CHOICE ANALYSIS:

A. Incorrect. The first part is correct. The second part is plausible because SDM verification is step 6 in the AP, but step 6 is skipped over when a rod has not dropped.

B. Correct. IAW TS 3.12.E, IRPI and DRPI may deviate by 24 steps for up to one hour each day. Since the deviation has existed for a max of 30 minutes, the conditions of the LCO are still met and will be for at least another half hour. SDM verification is not required IAW the procedure because the rod has not dropped.

C. Incorrect. See above.

D. Incorrect. See above.

REFERENCES TO BE PROVIDED TO THE APPLICANT: None

REFERENCES:

1. Tech Spec 3.12.E, ROD POSITION INDICATION SYSTEM AND BANK DEMAND POSITION INDICATION SYSTEM.

2. 0-AP-1.01, CONTROL ROD MISALIGNMENT, Rev. 20.

Answer: B

59. 015K2.01 1 Which one of the following completes the below statements?

(1) The normal power supply to NFI-NM-190A/B, the remote (local) Unit 1 Excore Neutron Flux Monitor (Excore Fission Chamber) used for Appendix R purposes, is (2) In the event of a loss of power on Unit 1, the power supply for NFI-NM-190A/B may be aligned to Unit 2 using a transfer switch located in A. (1) Vital Bus 1-I (2) the Unit 1 Cable Tray Room B. (1) Vital Bus 1-I (2) the Unit 2 Emergency Switchgear Room (ESGR)

C. (1) Vital Bus 1-II (2) the Unit 1 Cable Tray Room D. (1) Vital Bus 1-II (2) the Unit 2 Emergency Switchgear Room (ESGR)

K/A

015 Nuclear Instrumentation System 015K2.01 Knowledge of bus power supplies to the following: NIS channels, components, and interconnections.

(CFR 41.7) (RO - 3.3)

K/A Match Analysis This question matches the K/A statement by requiring the RO applicant in a straightforward fashion recall the Appendix R electrical distribution system that powers the remote indication Nuclear Instrument system.

Answer Choice Analysis A. INCORRECT. Part (1) of the distractor is correct; the normal power supply to the remote (Channel I) excore neutron flux monitor NFI-NM-190A/B is Vital Bus 1-I. Part (2) of the distractor is incorrect; the transfer switch is located in the Unit 2 ESGR. Part (2) is plausible because there is another transfer switch located in the Unit 1 Cable Tray room, which is used to transfer power to a LFFG portable generator (Large Fire or Flooding Generator).

B. CORRECT. See analysis of A above.

C. INCORRECT. Part (1) of this distractor is incorrect. However, Vital Bus 1-II is plausible because it is the normal power supply to the channel 2 excore neutron flux monitor, which is only read in the MCR (not remotely for Appendix R purposes). Part (2) is also incorrect, see analysis of A above.

D. INCORRECT. Part (1) is incorrect, part (2) correct.

Supporting References

1. Surry Lesson Plan ND-93.2-LP-5, EXCORE FISSION CHAMBER SYSTEM, rev.

5. Especially page ND-93.2-H/T-5.6, APPENDIX R POWER SUPPLY.

References Provided to Applicant None.

Answer: B

60. 029K3.01 1 Plant conditions:

- Unit 1 and Unit 2 are both in cold shutdown

- Containment Purge is in operation for both Units

- Containment purge vacuum breaker, located between the Unit 1 supply system penetrations valves, fails open Based on the above conditions, which one of the following states the effect on containment pressure?

A. Unit 1 containment pressure decreases.

Unit 2 containment pressure is unaffected.

B. Unit 1 containment pressure decreases.

Unit 2 containment pressure decreases.

C. Unit 1 containment pressure increases.

Unit 2 containment pressure is unaffected.

D. Unit 1 containment pressure increases.

Unit 2 containment pressure increases.

K/A Containment Purge: Knowledge of the effect that a loss or malfunction of the Containment Purge System will have on the following: Containment parameters.

K/A Match Analysis Requires knowledge of how system failure will affect containment pressure.

Answer Choice Analysis A. INCORRECT.

B. CORRECT. The result would be that less air would be going into containment, but the exhaust is still in operation.

C. INCORRECT.

D. INCORRECT. See above. Plausible because the valve is designed to raise pressure so that it is equal to atmospheric.

Supporting References ND-88.4-LP-6 Obj D, Revision 9 References Provided to Applicant None.

Licensee to verify this question based on what containment pressure is maintained at when containment purge is operating.

Answer: B

61. 034A2.01 1 Initial conditions:

• Unit 1 is in REFUELING

• Fuel off-loading is in progress.

Current conditions:

• A report is received that a spent fuel assembly being removed from the core has slipped from the manipulator crane and is lying on the upper core plate.

• Bubbles were seen escaping the dropped fuel assembly.

• No radiation alarms have been received from the radiation monitors associated with the manipulator crane or the refueling area.

• HP is monitoring the area, but report normal readings around the reactor cavity area.

Based on current conditions, which one of the following describes the actions required by 0-AP-22.00, Fuel Handling Abnormal Conditions?

A. Stop fuel handling operations and evacuate Containment. Fuel Building evacuation is not required.

B. Stop fuel handling operations, Notify Shift Supervision, OMOC and the Shift Technical Advisor. Area evacuations are not required at this time.

C. Stop fuel handling operations and evacuate Containment and the Fuel Building.

D. Stop fuel handling operations, Notify Shift Supervision, Fuel Performance Analysis and Health Physics. Area evacuations are not required at this time.

K/A Fuel Handling Equipment Ability to (a) predict the impacts of the following malfunctions or operations on the Fuel Handling System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Dropped fuel element.

(CFR: 41.5/43.5/45.3/45.13) (RO - 3.6)

K/A Match Analysis

The RO applicant is required recognize the impact from a dropped fuel assembly with normal radiation levels results in termination of fuel handling operations and evacuation of the containment based on the actions contained in 0-AP-22.00, Fuel Handling Abnormal Conditions.

Answer Choice Analysis AA. CORRECT. These are the actions that would be performed as a result of the dropped assembly. Fuel building would not require evacuation since the dropped assembly occurred inside containment.

B. INCORRECT. Plausible because with no detectable radiation the applicant may think that evacuation is not required. Following a dropped fuel assembly, step 2 of 0-AP-22.00 states Check local radiation levels - Normal. With normal radiation levels the user continues on to step 3 which directs the user to step 20 to make the listed notifications and then exit the procedure. However, the user would never get to step 2 because the RNO actions for step 1 (Check for fuel repairs) sends the user to step 4.0 which directs stopping fuel handling operations.

C. INCORRECT. Plausible because evacuation of both areas (step 5) is addressed by 0-AP-22.00. However, it is an OR statement rather than an AND statement.

D. INCORRECT. Plausible because with no detectable radiation the applicant may think that evacuation is not required. The second sentence is plausible because the precautions and limitations (4.3) section of 1-OP-FH-001, Controlling Procedure for Refueling, states that the Refueling SRO (Shift Supervision) and Fuel Performance Analysis will be notified immediately anytime fuel assembly damage is observed or expected. However, these positions are not listed in 0-AP-22.00 as requiring notification. In addition, the Health Physics department is notified only if abnormal radiation levels are observed.

Supporting References

1. Surry lesson plan ND-92.5-LP-7, "Refueling APs," rev. 13, Objective C, pp. 9-10.

2. 0-AP-22.00, Fueling Handling Abnormal Conditions, Rev. 23, pgs. 2 and 6.

References Provided to Applicant None.

Answer: A

62. 041K6.03 1

Unit 1 Initial Conditions:

• Power = 100%.

• G-A-6 ROD CONTROL SYSTEM URGENT FAILURE is in alarm.

• I&C has determined that the fault is in the logic cabinets and rod control is in AUTO to allow for troubleshoot the urgent failure.

Current Conditions:

• Main Turbine control valves unexpectedly throttle closed, resulting in a 12% load rejection over a 90 second period.

Based on the current conditions, which one of the following describes the effect on Reactor Coolant System (RCS) temperature with no operator action?

A. RCS temperature will change based on the higher Xenon concentration with no rod motion and no steam dump operation.

B. RCS temperature will be controlled entirely by control rods. No steam dumps will operate.

C. RCS temperature will be controlled entirely by steam dump operation. No control rods will move.

D. RCS temperature will be controlled with a combination of control rod motion and steam dump operation.

K/A 041 Steam Dump System (SDS)/Turbine Bypass Control 041K6.03 Knowledge of the effect of a loss or malfunction on the following will have on the Steam Dump System: Controller and Positioners, including ICS, S/G, CRDS.

(CFR 41.7 / 45.7) (RO - 2.7)

K/A Match Analysis This question matches the K/A statement by requiring the RO applicant to correctly assess the effect of a rod control malfunction (CRDS) on the Steam Dump System and the control of RCS temperature, given an operationally valid/plausible situation.

Answer Choice Analysis A. INCORRECT. Steam dumps will operate as stated in correct answer. Plausible because rods will not operate and applicant may not make the connection that steam

dumps are armed with a demand signal.

B. INCORRECT. Control Rods will not move due to the urgent failure. Plausible if applicant does not know that the alarm is indication that rods will not move because rods are capable of handling 10% load rejects without the help of the steam dumps.

C. CORRECT. Rods will not move due to the urgent failure. Steam dumps will arm with a 10% rejection in less than 2 minutes. Therefore, with an arming signal and a Tave-Tref deviation (hence a demand signal), steam dumps will open.

D. INCORRECT. Rods will not move due to the urgent failure. Plausible if applicant does not know that the alarm is indication that rods will not move.

Supporting References

1. Slightly modified from Vogtle 2005-301 ILO exam question 041K6.10, still counting as BANK question.

2. Surry Lesson Plan ND-93.3-LP-9, STEAM DUMPS, rev. 13.

3. Surry Lesson Plan ND-93.3-LP-3, ROD CONTROL SYSTEM, rev. 19.

References Provided to Applicant None.

Answer: C

63. 045A1.05 1 Unit 1 initial conditions:

Reactor shutdown in progress Reactor power = 8%

Turbine Trip occurs Main Steam Dump Valves fail closed Based on the above conditions, which one of the following states (1) if the Pzr PORV will open during the initial transient (1st minute) and (2) the Tcold at which the RCS will stabilize?

A. (1) Yes (2) 550 oF B. (1) Yes (2) 556 oF

C. (1) No (2) 550 oF D. (1) No (2) 556 oF K/A Main Turbine Generator: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MT/G system controls including:

Expected response of primary plant parameters (temperature and pressure) following T/G trip.

K/A Match Analysis Requires applicant to know how primary system pressure and temperature will respond to a turbine trip.

Answer Choice Analysis A. Incorrect: 1st part is plausible in that the normal secondary heat sink has been lost with the reactor still at power. 2nd part is correct.

B. Incorrect: 1st part is plausible in that the normal secondary heat sink has been lost with the reactor still at power. With the main steam dumps failed closed, RCS Tcold will stabilize at the saturation temperature of the SGs. The SG pressure will be maintained by the SG PORVs at1035 psig --- 1050 psia --- 550 0F Tsat.

C. Correct: The Pzr spary valves are designed to prevent lifting the Pzr Porv during a 10% step decrease in power. The 3 SG PORVs have a capacity of ~ 1.1 E6 lbm/hr.

With the main steam dumps failed closed, RCS Tcold will stabilize at the saturation temperature of the SGs. The SG pressure will be maintained by the SG PORVs at1035 psig --- 1050 psia --- 550 0F Tsat.

D. Incorrect: 1st part is correct. 2nd part is plausible because if the steam demand were to exceed the capacity of the SG PORVs, it could be correct based on the lowest SG Safety setpt of 1085 psig --- 1100 psia --- 556 0F Tsat.

Supporting References ND-89.1-LP-2 ND-88.1-LP-3 References Provided to Applicant none Licensee to verify RCS temperature and Pressurizer abbreviation

Answer: C

64. 071K4.01 1 Which one of the following corresponds to the pressure setting of the relief valve on the inner tank of the Waste Gas Decay Tanks?

A. 75 psig.

B. 100 psig.

C. 135 psig.

D. 150 psig.

K/A Waste Gas Disposal Knowledge of design feature(s) and/or interlock(s) which provide for the following:

Pressure capability of the waste gas decay tank.

(CFR: 41.7) (RO - 2.6)

K/A Match Analysis The RO applicant is required to know the pressure the inner tank is capable of withstanding before its contents are released to the surrounding environment via the relief valve.

Answer Choice Analysis A. INCORRECT. Plausible because this is the alarm point for annunciator 0-WD-B9, WASTE GAS DECAY TKS RELIEF VVS HI PRESS. The alarm is sensing a high pressure condition between the relief valve and the downstream rupture disk.

B. INCORRECT. Plausible because this is the pressure setting for the relief valve on the outer tank of the Waste Gas Decay Tanks.

C. INCORRECT. Plausible because this is the alarm point for annunciator 0-WD-C10, INNER WASTE GAS DECAY TKS HI PRESS.

D. CORRECT. The relief valve on the inner tank of the Waste Gas Decay Tanks is set to lift at 150 psig.

Supporting References

1. Surry lesson plan ND-92.4-LP-1, "Gaseous and Liquid Waste Processing Systems,"

rev. 13, Obj. B, pp. 13-14.

References Provided to Applicant None.

Answer: D

65. 072G2.2.40 1 Unit 2 is about to begin refueling operations (movement of irradiated fuel and water level is >23 feet). Per T.S. 3.10, Refueling, which ONE of the following will prevent refueling operations from commencing?

A. The 1A RHR pump was declared inoperable 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> ago, both RHR HXs are available.

B. The Manipulator Crane Area Rad. Monitors have failed and containment purge is isolated.

C. The inner door of the personnel airlock cannot be closed.

D. RCS boron concentration is 2360 ppm as verified by a sample taken 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ago.

K/A Area Radiation Monitoring / Area Radiation Monitoring: Ability to apply Technical Specifications for a system.

K/A Match Analysis Requires knowledge of TS and how to apply statements of applicability.

Answer Choice Analysis A. INCORRECT. T.S. 3.10 only requires one RHR pump and heat exchanger to be OPERABLE as long as water level is >23 feet.

B. CORRECT. T.S. 3.10 states the Manipulator Crane Area Rad. Monitors must be OPERABLE and continuously monitored.

C. INCORRECT. T. S. 3.10 states the equipment access hatch and at least one door in the personnel airlock shall be capable of being closed.

D. INCORRECT. T.S. 3.10 requires boron to be greater than 2300 ppm and sampled every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Supporting References T.S. 3.10.A.4 ND-92.5-LP-1, Obj. D References Provided to Applicant None.

Answer: B

66. G2.1.25 1 Unit 1 Current conditions:

• Reactor has been shutdown for 5 days.

• RCS water level is being maintained at 7.0 feet as indicated 1-RC-LR-105.

• The B and C loops are isolated.

• Steam generator (SG) manways have been removed in preparation for SG tube plugging efforts.

• The reactor vessel head is tensioned.

• A RHR pump is in operation with oscillating amperage indications

• Flow indication on 1-RH-FI-1605 is oscillating between 3,300 and 3,700 gpm.

Based on the current conditions, which one of the following describes the proper adjustments of RCS level and RHR flow in accordance with 1-AP-27.00, Loss of Decay Heat Removal Capability?

(REFERENCE PROVIDED)

A. Raise RCS level to 9.0 feet as indicated on 1-RC-LR-105 and reduce flow to 3,100 gpm.

B. Maintain current RCS level as indicated on 1-RC-LR-105 and reduce flow to 1,500 gpm.

C. Maintain current RCS level as indicated on 1-RC-LR-105 and reduce flow to 3,100 gpm.

D. Raise RCS level to 9.0 feet as indicated on 1-RC-LR-105 and stabilize flow at 3,500

gpm.

K/A Plant Wide Generics Ability to obtain and interpret station reference materials such as graphs, monographs, and tables, which contain performance data.

(CFR: 41.10/43.5/45.12) (RO - 3.9)

K/A Match Analysis The RO applicant must recognize that the RHR pump is vortexing and use the graphs on Attachment 1, RHR Flow Requirements Versus Time after Shutdown and , Minimum RCS Level Versus RHR Flow (1-RC-LR-105) to determine the RHR flow and RCS level that will stop pump vortexing.

Answer Choice Analysis A. CORRECT. Per Attachment 1, a flow rate of 3,100 gpm satisfies the minimum RHR flow for the five days (120 hrs) the reactor has been shutdown. Per Attachment 3, a flow rate of 3,100 gpm requires a minimum RCS level of ~8.6 feet. With RCS level at 9.0 feet and RHR flow at 3,100 gpm ends up in the acceptable region of the graph in .

B. INCORRECT. Plausible because the conditions of RCS level at 7.0 feet and RHR flow at 1,500 gpm ends up in the acceptable region of the Attachment 3 graph.

However, the RHR flow is not adequate to satisfy the minimum flow requirements for for the given shutdown period of five days. .

C. INCORRECT. Plausible because a flow 3,100 gpm will satisfy the flow requirements for Attachment 1 for the given shutdown period of five days. However, the RCS level of 7.0 feet does not provide adequate NPSH at 3,100 gpm flow to prevent vortexing per Attachment 3.

D. INCORRECT. Plausible because a flow 3,500 gpm will satisfy the flow requirements for Attachment 1 for the given shutdown period of five days. However, the RCS level of 9.0 feet does not provide adequate NPSH at 3,500 gpm flow to prevent vortexing per Attachment 3.

Supporting References

1. Surry lesson plan ND-88.2-LP-3, "Draindown and Mid-Loop Operations," rev. 16, Objective C, pp. 20-23.

2. 1-AP-27.00, Loss of Decay Heat Removal Capability, Rev. 20

3. This question was used on the Surry 2004-301 NRC exam. Modified conditions slightly and re-organized distractors, but question is essentially the same.

References Provided to Applicant Attachments 1 and 3 of 1-AP-27.00 Answer: A

67. G2.1.37 1 Which one of the following statements lists the required items to be in a Reactivity Plan for a reactor startup in accordance with GOP-1.4, UNIT STARTUP, HSD TO 2%

REACTOR POWER?

A. Delta flux control, expected xenon transient, and recommendations for rod height and/or RCS boron adjustments.

B. Limitations on startup rate, expected xenon transient, and recommendations for rod height and/or RCS boron adjustments.

C. Delta flux control, expected xenon transient, and RCS temperature control.

D. Delta flux control, expected source range counts at the doubling points, and recommendations for rod height and/or RCS boron adjustments.

K/A G2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management?

KA MATCH ANALYSIS:

The above question requires knowledge of the components of a Reactivity Plan, which is testing knowledge of reactivity management guidelines.

ANSWER CHOICE ANALYSIS:

A. Correct. See list from GOP-1.4 B. Incorrect. The only incorrect item is the limitations on SUR. Plausible because there are limits on SUR that likely would be included in a reactivity brief.

C. Incorrect. The only incorrect item is RCS temperature. Plausible because there are limitations on RCS temperature that likely would be included in a reactivity brief.

D. Incorrect. The only incorrect item is SR counts at doubling points. This too is plausible because it is a likely component for a reactivity brief for performing this procedure.

REEFERENCES TO BE SUPPLIED TO APPLICANT: None

Memory Level / Lower Cog / Fundamental LOD = 2 New Question REFERENCS:

GOP-1.4, UNIT STARTUP, HSD TO 2% REACTOR POWER, Rev 34.

REFERENCES TO BE SUPPLIED TO APPLICANT: None Answer: A

68. G2.2.13 1 Which one of the following completes the statement listed below, in accordance with OP-AA-200, Equipment Clearance?

IF maintenance activities are to be performed on a (1) _____________________ that would normally be tagged OPEN, THEN ENTER the component on the Tagging Record to show the initial and final positions to maintain status control as a (2)

A. (1) breaker (2) Operating Permit (blue lock), and only one Tagout Holder for each tag-out is allowed.

B. (1) breaker (2) No Tag item, and only one Tagout Holder for each tag-out is allowed.

C. (1) vent or drain valve (2) Operating Permit (blue lock), and more than one Tagout Holder for each tag-out is allowed.

D. (1) vent or drain valve (2) No Tag item, and more than one Tagout Holder for each tag-out is allowed.

K/A Generic: Equipment Control G2.3.13 Knowledge of tagging and clearance procedures.

(CFR 41.10 / 45.13) (RO - 4.1)

K/A Match Analysis This question matches the K/A statement by requiring the RO applicant to correctly apply knowledge of the Dominion tagging and clearance procedure in order to properly tag (or, in this case, no tag) a vent valve that needs to have maintenance performed on the valve.

Answer Choice Analysis A. INCORRECT. Both part (1) and part (2) are incorrect. OP-AA-200 step 3.4.1.c.

states IF maintenance activities are to be performed on a vent or a drain that would normally be tagged open, THEN ENTER the component on the Tagging Record to show the initial and final positions to maintain status control as a No Tag item. It is plausible that maintenance may need to be performed on a breaker that is already part of a tagging order; in this case, OP-AA-200 section 3.3.7.e states that the Danger Tag is moved to the breaker cubicle face or to the breaker door, then returned to the racking device or appropriate electrical isolation location. OP-AA-200 section 3.5 describes the use of Operating Permits (blue locks), which allows the Permit Holder (Tagout Holder) the authorization to OPERATE (but not do maintenance on) a designated component.

Therefore, Operations Permit is also a plausible, but incorrect, distractor.

B. INCORRECT. See analysis of A above.

C. INCORRECT. See analysis of A. above.

D. CORRECT. See analysis of A. above.

Supporting References

1. Dominion Nuclear Fleet Administrative Procedure OP-AA-200, Equipment Clearance, rev. 4.

References Provided to Applicant None.

Answer: D

69. G2.2.35 1 Which one of the following describes the Technical Specification definitions of (1) COLD SHUTDOWN, and (2) REFUELING SHUTDOWN?

A. (1) At least 1 % delta-k/k & Tave less than or equal to 200 oF

(2) At least 5 % delta-k/k, Tave less than or equal to 140 oF with fuel scheduled to be moved.

B. (1) At least 1 % delta-k/k & Tave less than or equal to200oF (2) Reactor vessel head unbolted.

C. (1) At least 1.77 % delta-k/k & Tave less than or equal to 200oF (2) At least 5 % delta-k/k, Tave less than or equal to 140 F with fuel scheduled to be moved.

D. (1) At least 1.77 % delta-k/k & Tave less than or equal to 200oF (2) Reactor vessel head unbolted.

K/A G2.2.35 Ability to determine TS mode of operation.

K/A Match Analysis Q tests memory of TS Mode Definitions.

Answer Choice Analysis A. Correct. See Tech Spec Definitions B. Incorrect. Plausibility of part 2 exists with the potential confusion with REFUELING OPERATIONS.

C. Incorrect. Plausibility of part 1 exists because 1.77% is the SDM requirement for INTERMEDIATE SHUTDOWN.

D. Incorrect.

Supporting References Tech Specs - Section 1 Definitions References Provided to Applicant None.

Answer: A

70. G2.2.39 1 Current Unit 1 Conditions:

- Plant is in COLD SHUTDOWN

- CH-223 (facility to provide name of component) has been opened for makeup activities Which one of the following (1) describes the maximum allowable Tech Spec time for

1-CH-223 to be open, and (2) whether sealing the valve after closure will comply with the conditions of the Tech Spec LCO 3.2, Chemical and Volume Control System?

A. 15 minutes Sealing the valve after closure will comply with the Tech Spec LCO.

B. 15 minutes Sealing the valve after closure will NOT comply with the Tech Spec LCO.

C. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Sealing the valve after closure will comply with the Tech Spec LCO.

D. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Sealing the valve after closure will NOT comply with the Tech Spec LCO.

K/A Knowledge of less than one hour Technical Specification action statements for systems.

K/A Match Analysis Applicant must have knowledge of a 15 minute LCO action statement in order to arrive at the correct answer.

Answer Choice Analysis A. CORRECT. Tech Spec 3.2.E states that this and other valves may be opened for 15 minutes for makeup and/or planned dilution activities. Upon closing Tech Specs requires them to be locked, sealed, or otherwise secured.

B. INCORRECT. The second part is incorrect because the Tech Specs state that sealing the valve is acceptable. Plausible because the tech specs also state that locking the valve is acceptable; therefore, an applicant could have a credible misconception that the sealing the valve is not enough and think that the valve must be locked.

C. INCORRECT. See above.

D. INCORRECT. See above.

Supporting References Modified from NRC 2004-301 exam.

Tech Spec 3.2.E

References Provided to Applicant None.

Answer: A

71. G2.3.12 1 Current conditions on Unit 1:

• Core loading is in progress.

• Both loops of RHR are operable with B RHR loop providing shutdown cooling.

• Reactor cavity level is dropping.

Based on the current conditions, which one of the following describes :

(1) the minimum level above the reactor vessel flange that must be maintained in the reactor cavity during REFUELING OPERATIONS in accordance with Technical Specifications and (2) the reason for maintaining the reactor cavity level above this limit?

A. (1) 26 feet (2) To ensure the fuel remains cooled for the period of time it takes to place A

RHR loop in-service in the event of a failure on B RHR loop.

B. (1) 26 feet (2) To ensure 99% of the radioactive iodine gas is removed in the event an irradiated fuel assembly ruptures.

C. (1) 23 feet (2) To ensure the fuel remains cooled for the period of time it takes to place A

RHR loop in-service in the event of a failure on B RHR loop.

D. (1) 23 feet (2) To ensure 99% of the radioactive iodine gas is removed in the event an irradiated fuel assembly ruptures.

K/A Reactor Trip 007EK3.01 Knowledge of the reasons for the following as they apply to a reactor trip:

Actions contained in EOP for reactor trip.

(CFR 41.5 / 41.10 / 45.6 / 45.13) (RO - 4.0)

K/A Match Analysis The question requires the applicant to understand a licensed operators fuel handling responsibilities including knowledge of those plant conditions that would require stopping fuel movement.

Answer Choice Analysis A. INCORRECT. Plausible because 26 is the low level alarm point that is established during REFUELING OPERATIONS. The second half is plausible because a reactor cavity level of 23 feet is the point where two RHR loops are required to be operable.

The reason for the level requirement is because with only one RHR loop operable the inventory in the reactor cavity would allow time to initiate emergency procedures not to swap RHR loops.

B. INCORRECT. Plausible because 26 is the low level alarm point that is established during REFUELING OPERATIONS. The second half is the correct reason for the limit.

C. INCORRECT. Plausible because the first half of the answer is the correct limit.

The second half is plausible because a reactor cavity level of 23 feet is the point where two RHR loops are required to be operable. The reason for the level requirement is because with only one RHR loop operable the inventory in the reactor cavity would allow time to initiate emergency procedures not to swap RHR loops.

D. CORRECT. T.S. 3.10.A.6 requires a minimum 23 feet above the reactor vessel flange in the reactor cavity and the reason is to ensure 99% of the iodine gas is removed in the event of a ruptured irradiated fuel assembly Supporting References

1. Surry lesson plan, ND-92.5-LP-1, Refueling Overview, Rev. 17, Obj. D, pg.17

2. T.S. 3.10.A.7, pg 3.10-3 References Provided to Applicant None.

Answer: D

72. G2.3.7 1 Which of the following activities requires a Specific Radiation Work Permit (RWP), in

accordance with VPAP-2101, Radiation Protection Program?

A. An entry into a posted Airborne Radioactivity Area for work in an airborne radiation field of 100 DAC.

B. An entry into a posted Locked High Radiation Area (LHRA) with work to be performed in a radiation field of 100 mrem/hr.

C. An entry into a posted High Radiation Area (HRA) with a work team of three (3) individuals that will result in an estimated exposure of 100 mrem for the team.

D. An entry for a Planned Special Exposure (PSE) for work that will result in exceeding the annual regulatory limit by a factor of 100%.

K/A Generic: Radiation Control G2.3.7 Ability to comply with radiation work permit requirements during normal or abnormal conditions.

(CFR 41.12 / 45.10) (RO - 3.5)

K/A Match Analysis This question matches the K/A statement by requiring the RO applicant to correctly remember the requirements for Specific RWPs, given plausible operationally valid scenarios.

Answer Choice Analysis A. INCORRECT. Entry into an airborne activity area requires additional internal monitoring, but does not require a Specific RWP.

B. CORRECT. VPAP-2101 section 6.8.5 states : Use of General RWPs is prohibited under either of the following conditions: - Entry into a Locked High Radiation Area or Very High Radiation Area, - Expected exposure is greater than 100 mrem per individual per entry.

C. INCORRECT. See analysis of B. above. A specific RWP is required with expected exposure is greater than 100 mrem per individual per entry.

D. INCORRECT. Although other controls are required for a PSE, a specific RWP is not required. As a note, exceeding the annual limit by 100% is not allowed, but was included in the distractor stem for maximum plausibility among the distractors.

Supporting References

1. Dominion Station Administrative Procedure VPAP-2101, Radiation Protection Program, rev. 34.

2. Question G2.3.7 from Farley Exam in 1998.

References Provided to Applicant None.

Answer: B

73. G2.4.19 1 Which one of the following states (1) the meaning of an

• when displayed by a step number in an EOP, and (2) whether a CAUTION statement or a NOTE is used in the EOPs to advise on actions or transitions which may become necessary depending on changes in plant conditions?

A. (1) Immediate Operator Action (2) CAUTION statements are used to advise on actions or transitions which may become necessary depending on changes in plant conditions.

B. (1) Continuous Action (2) CAUTION statements are used to advise on actions or transitions which may become necessary depending on changes in plant conditions.

C. (1) Immediate Operator Action (2) NOTES are used to advise on actions or transitions which may become necessary depending on changes in plant conditions.

D. (1) Continuous Action (2) NOTES are used to advise on actions or transitions which may become necessary depending on changes in plant conditions.

K/A Knowledge of EOP layout, symbols, and icons.

K/A Match Analysis Question requires knowledge of symbols used in the EOPs.

Answer Choice Analysis A. INCORRECT. Immediate action is plausible because [ ] are used in the EOPs to designate immediate actions.

B. CORRECT. See pages 6 and 17 of listed reference.

C. INCORRECT. Plausible because CAUTIONS are also used for advising of potential hazards to personnel or equipment. NOTES are used for administrative or advisory information which support operator actions. It is credible for an operator to confuse the two.

D. INCORRECT. See above.

Supporting References ND-95.3-LP-2, EMERGENCY PROCEDURE WRITER'S FORMAT, Rev 13.

References Provided to Applicant None.

Answer: B

74. G2.4.42 1 Initial Conditions:

• A large group of armed hostile intruders has gained access to multiple vital areas of the Surry Power Station.

Current Conditions:

• The Shift Manager has declared an emergency, and determined that the Local Emergency Operations Facility (LEOF) is unavailable due to the ongoing gun battle occurring on the station.

Based on the current conditions, which one of the following is (1) the backup facility for the LEOF; AND (2) personnel Accountability must be complete within __________

following a declaration of ALERT, SITE AREA EMERGENCY, or GENERAL EMERGENCY, in accordance with the Surry Emergency Plan (SEP)?

A. (1) Corporate Emergency Response Center (CERC)

(2) 60 minutes B. (1) Central Emergency Operations Facility (CEOF)

(2) 30 minutes C. (1) Corporate Emergency Response Center (CERC)

(2) 30 minutes D. (1) Central Emergency Operations Facility (CEOF)

(2) 60 minutes K/A Generic: Emergency Procedures/Emergency Plan G2.4.42 Knowledge of emergency response facilities.

(CFR 41.10 / 45.11) (RO - 2.6)

K/A Match Analysis This question matches the K/A statement by requiring the RO applicant to correctly remember the alternate sites for the EOF and the LMC, given a plausible operational scenario that would necessitate the use of E-plan response facilities offsite to the Surry plant. Applicants must also recall an important metric for Personnel Accountability.

Answer Choice Analysis A. INCORRECT. Part (1) of the distractor is incorrect, part (2) is also incorrect. The Surry Emergency Plan states Should the LEOF become unavailable during an emergency the responsibilities assigned to the LEOF will be transferred to the backup facility known as the Central Emergency Operations Facility. Corporate Emergency Response Center (CERC) is plausible because it is located at the same site (Innsbruck) as the CEOF. Part (2) is incorrect, because the Surry lesson plan for Emergency Plan Overview states: "Accountability must be complete within 30 minutes following declaration of Alert Site Area Emergency or General Emergency." 60 minutes is plausible because it is the normal time allowed to fully staff the EOF/TSC etc.

B. CORRECT. See analysis of A. above.

C. INCORRECT. Part (1) incorrect, part (2) correct. See analysis of A. above.

D. INCORRECT. Part (1) is correct, part (2) incorrect. See above descriptions.

Supporting References

1. Surry Emergency Plan Procedure, Surry Power Station Emergency Plan, rev. 54 dtd 12/30/2008. Especially chapter 7.

2. Surry Lesson Plan ND-95.5-LP-1, "EMERGENCY PLAN OVERVIEW," rev. 8 dtd 07/29/08. See p. 18.

References Provided to Applicant None.

Answer: B

75. G2.4.45 1 Unit 1 current conditions:

• The reactor is operating at 100% power.

• Condenser vacuum is 26 Hg and lowering at 0.1 Hg/minute.

• Steam generator levels are 45% and lowering.

• Several annunciators are lit, including.

• 1H-G8, FW PP DISCH HDR LO PRESS

• 1J-G4, CN PPS DISCH HDR LO PRESS

• 1C-A1, RCP 1A CC RETURN LO FLOW

• 1C-B1, RCP 1B CC RETURN LO FLOW

• 1C-C1, RCP 1C CC RETURN LO FLOW Which one of the following correctly identifies the procedure required to be entered and the event that is consistent with the above alarms and conditions?

A. 1-AP-9.02, Loss of RCP Seal Cooling.

B. 1-AP-10.05, Loss of Semi-Vital Bus.

C. 1-AP-21.00, Loss of Main Feedwater Flow.

D. 1-AP-10.01, Loss of Vital Bus I.

K/A Plant Wide Generics Ability to prioritize and interpret the significance of each annunciator or alarm.

(CFR: 41.10/43.5/45.3/45.12) (RO - 4.1)

K/A Match Analysis The question requires the RO applicant to prioritize the listed alarms by significance

based on the plant conditions and recognize that a failure of the semi-vital bus has occurred.

Answer Choice Analysis A. INCORRECT. Plausible because the alarms would appear to indicate a loss of CC cooling to the Thermal Barrier Heat Exchangers on the RCPs However, this AP is only entered if CC Cooling is lost along with a loss of RCP seal injection or elevated RCP seal/ bearing temperatures.

B. CORRECT. All the conditions and alarms listed in the stem of the question are consistent with a loss of the semi-vital bus. The feedwater and condensate recirculation valves fail open on a loss of the semi-vital bus which is consistent with low discharge pressure alarms for condensate and feedwater. Condenser vacuum would degrade since auxiliary steam would isolate to the air ejectors on a loss of the semi-vital bus.

C. INCORRECT. Plausible because some of the alarms and conditions indicate a loss of feedwater has occurred. While a loss of the condensate system would result in the low discharge pressure alarms for condensate and feedwater, it would not explain the alarms on the CC cooling to the RCPs or degrading condenser vacuum.

D. INCORRECT. Plausible because CC cooling flow to the thermal barrier heat exchangers would be effected (closes the RCP thermal barrier CC Cooling return valve CC-TV-140B) on a loss of Vital Bus 1. However, it would not explain the feedwater/condensate low discharge pressure alarms or the degraded condenser vacuum condition.

Supporting References

1. Surry lesson plan, ND-90.3-LP-5, Vital and Semi-Vital Bus Distribution, Rev. 17, Obj. E, pp. 15-16 and 21-22.

References Provided to Applicant None.

Answer: B