ML102780354
| ML102780354 | |
| 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: ML102780354 (53) | |
Text
1. 000000000015AG2.4.41 1 Unit 1 Initial Conditions:
- Reactor Power = 100%
Unit 1 Current Conditions:
E-0, Reactor Trip or Safety Injection has been initiated
- RCP-1A seal failure exists and the pump has been secured
- RCS leakage is 200 gpm
- CETs = 550oF
- Containment pressure = 20 psia
- Subcooled margin = 50oF
- 2 RCPs are operating
- RVLIS dynamic head range = 30%
Based on the current conditions, which one of the following correctly states: (1) the EAL classification required to be made by the shift manager, and (2) the maximum time for notification of the NRC after the declaration is made, in accordance with EPIP-2.02, Notification of NRC, ?
(REFERENCE PROVIDED)
A. (1) Alert (2) 15 minutes B. (1) Alert (2) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> C. (1) Site Area Emergency (2) 15 minutes D. (1) Site Area Emergency (2) 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> K/A RCP Malfunctions Knowledge of emergency action level thresholds and classifications K/A Match Analysis The question requires the applicant to apply the EAL table, which will be provided to the applicant, to determine the correct classification. The applicant must determine that an ORANGE path exists on Core Cooling, thereby indicating a potential loss of the fuel
clad barrier and a loss of the RCS barrier.
SRO Guidance Analysis Classification of an event is strictly an SED function, of which all SRO applicants are responsible. The EAL table will be provided, but the question is not a direct lookup because the applicant must recognize that an ORANGE path exist on core cooling, thus giving a potential loss of fuel cladding. The elevated RCS leakage also indicates a loss of the RCS barrier. This loss of two barriers creates the SAE classification.
Answer Choice Analysis A. INCORRECT. Alert is plausible because if the applicant does not diagnose the fuel cladding potential loss, then ALERT would be correct. Fifteen minutes is plausible because the STATE and LOCALs are required to be notified within 15 minutes - the plausibility plays off the applicant potentially confusing the 15 and 60 minute requirements.
B. INCORRECT. See analysis on A and D.
C. INCORRECT. See analysis on A and D.
D. CORRECT. The SFST for core cooling has been attached with the flow path highlighted based on parameters provided in the stem. A SAE exists due to the loss of the RCS barrier and the potential loss of the fuel cladding. The NRC is required to be notified within one hour of event declaration.
Supporting References F-2, CORE COOLING, rev. 1 EAL Matrix EPIP-1.01, Emergency Manager Controlling Procedure, Rev. 52.
EPIP-2.02, Notification of NRC, Rev. 21 References Provided to Applicant None.
Answer: D
- 2. 000000000022AG2.2.25 1 Current conditions:
- Unit 1 is at 100% power.
- Unit 2 is in INTERMEDIATE SHUTDOWN with the following charging system Alignment:
- Charging pumps 2B and 2C are inoperable
- Unit 2 RWST is the only source of supply to the charging system.
- A leak has been found on 2-CH-447, Charging Pumps Cross-connect to Unit 1 Isolation Valve, and has been isolated for repairs In accordance with Technical Specifications, which one of the following identifies:
(1) the status of LCO 3.2, Chemical and Volume Control for both units, and (2)the basis for the LCO requirement?
Actions of LCO 3.2 are required on (1) due to the potential inability to (2) .
A. (1) Unit 1 but NOT Unit 2 (2) bring the plant to COLD SHUTDOWN conditions during specific postulated fire scenarios.
B. (1) Unit 1 but NOT Unit 2 (2) maintain a stable RCS makeup flowpath during specific postulated seismic scenarios.
C. (1) Unit 1 and Unit 2 (2) bring the plant to COLD SHUTDOWN conditions during specific postulated fire scenarios.
D. (1) Unit 1 and Unit 2.
(2) maintain a stable RCS makeup flowpath during specific postulated seismic scenarios.
K/A Loss of Reactor Coolant Makeup Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.
(CFR: 41.5/41.7/43.2) (SRO - 4.2)
Question History: New LOK: C/A LOD: 3 K/A Match Analysis
The SRO applicant must determine whether an LCO has been entered and the basis for the LCO to arrive at the correct answer.
SRO-Only Analysis In addition to knowing the basis for the LCO 3.2 concerning the availability of the opposite units charging system, the applicant must also evaluate the plant conditions, determine the LCOs applicability for both units, determine whether the requirements for available are met and apply the Technical Specification basis behind LCO 3.2. The question requires a more in-depth understanding of the specific LCO than would be expected for an RO applicant.
Answer Choice Analysis A. CORRECT. LCO 3.2 must be entered on Unit 1 because the ability to provide a source for boron injection from the opposite unit is not available. The bases states, In the event the operating units charging pumps become inoperable, this permits the opposite units charging pump to be used to bring the disabled unit to COLD SHUTDOWN conditions. The requirement is only applicable when the reactor is critical and is due to the potential of various fire scenarios taking out all charging pumps in a single unit.
B. INCORRECT. The first half of the response is correct. The second half is plausible because this is the basis behind FCA-14.00, Establish Stable CS Makeup Flowpaths.
However, this is based on a postulated fire scenario not a seismic scenario.
C. INCORRECT. Plausible because loss of the cross-connect line would result in entry into LCO 3.2 for Unit 1. It is plausible that entry into LCO 3.2 for Unit 2 would be required as well since the basis of the action is to bring the unit to COLD SHUTDOWN conditions and Unit 2 is still in INTERMEDIATE SHUTDOWN. However, the LCO requirement is only applicable when the reactor is critical. The second half of the response contains the correct basis for the requirement.
D. INCORRECT. Plausible because loss of the cross-connect line would result in entry into LCO 3.2 for Unit 1. It is plausible that entry into LCO 3.2 for Unit 2 would be required as well since the basis of the action is to bring the unit to COLD SHUTDOWN conditions and Unit 2 is still in INTERMEDIATE SHUTDOWN. The second half is plausible because this is the basis behind FCA-14.00, Establish Stable CS Makeup Flowpaths. However, this is based on a postulated fire scenario not a seismic scenario.
Supporting References
- 1. Surry lesson plan ND-88-3-LP-2, Charging and Letdown, Rev. 15, Obj. D, pg. 28.
- 2. Surry lesson plan ND-95.6-LP-2, Other Supporting FCAs, Rev. 12, Obj. E, pg. 11.
- 3. Technical Specifications, 3.2, Amdt. 199, pg 3.2-2.
References Provided to Applicant None Answer: A
- 3. 000000000027AG2.4.8 1 Unit 1 Initial Conditions:
- 100% Power.
- Multiple lightning strikes cause the following:
- PT-445 fails HIGH.
- Automatic Main Generator trip.
- While performing immediate operator actions of 1-E-0, REACTOR TRIP OR SAFETY INJECTION, the Reactor Operator notes the following:
- PT-444 = 1910 psig and INCREASING.
- PT-455 = 1930 psig and INCREASING.
- PT-456 = 1920 psig and INCREASING.
Current Conditions:
- Operators have transitioned to 1-ES-0.1, REACTOR TRIP RESPONSE, and are at the step to CHECK PRZR PRESSURE CONTROL.
- No adjustments were made to any PRZR Pressure Control component after the reactor trip.
- The Reactor Operator notes the following:
- PT-444 = 1990 psig and DECREASING.
- PT-455 = 2010 psig and DECREASING.
- PT-456 = 2000 psig and DECREASING.
Based upon the current conditions, (1) what is the NEXT required operator action to stop pressure from lowering, AND (2) what procedure(s) is/are required to be performed in parallel with 1-ES-0.1 to restore normal pressure control at 2235 psig?
A. (1) CLOSE the open PRZR Spray Valves.
(2) Perform 1-AP-31, INCREASING OR DECREASING RCS PRESSURE, to address BOTH the failed spray valve(s) AND the PT-445 failure. Entry may also be made to 0-AP-53.00, "LOSS OF VITAL INSTRUMENTATION/CONTROLS," which will direct a transition to 1-AP-31.
B. (1) CLOSE the open PRZR PORV(s) or associated block valve(s).
(2) Perform 1-AP-31 to address the PT-445 failure ONLY. Entry may also be made to 0-AP-53.00, which will direct a transition to 1-AP-31.
C. (1) CLOSE the open PRZR Spray Valves.
(2) Perform 0-AP-53.00 to address BOTH the failed spray valve(s) AND the PT-445 failure. A transition to 1-AP-31 will NOT be required.
D. (1) CLOSE the open PRZR PORV(s) or associated block valve(s).
(2) Perform 0-AP-53.00 to address the PT-445 failure ONLY. A transition to 1-AP-31 will NOT be required.
K/A Pressurizer Pressure Control System Malfunction 027AG2.4.8 Pressurizer Pressure Control System Malfunction: Knowledge of how abnormal operating procedures are used in conjunction with EOPs.
(CFR 41.10 / 43.5/ 45.13) (SRO - 4.5)
K/A Match Analysis This question matches the K/A statement by requiring the SRO applicant to correctly analyze a credible operational situation that involves a pressurizer pressure system component failure during EOP usage (1-ES-0.1), and to determine the correct procedural flow path to mitigate the situation.
SRO Guidance Analysis The question author intends to link this question to 10CFR55.43(b)(5) - assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. In accordance with rev. 1 of the SRO-Only Guidance, Figure 2, the key assessment for this question is that arriving at the correct answer requires Assessing plant conditions (normal, abnormal, or emergency) and then prescribing a procedure or section of a procedure to mitigate, recover, or with which to proceed.
A summary of the screening process is as follows:
- Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, location, etc.
Answer: NO
- Can the question be answered solely by knowing immediate operator actions?
Answer: NO
- Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?
Answer: NO
- Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?
Answer: NO
- Does the question require one or more of the following? [] Assessing plant conditions (normal, abnormal, or emergency) and then prescribing a procedure or section of a procedure to mitigate, recover, or with which to proceed Answer: YES - may be considered SRO-only.
Answer Choice Analysis A. INCORRECT. As detailed below, the plant is responding as expected for the PT-445 failure with no other operator actions. This choice is plausible if the applicant does not remember the PORV interlock at 2000 psig, and plausibility for this distractor is further enhanced by the correct procedural flow path answer choice for (2). IF PT-444 had failed high, both the PORVs and the spray valves would be open. Plausibility for spray valve failure being the cause of the lowering pressure is enhanced by the fact that verifying spray valves closed is 1-ES-0.1 step 6. b) RNO 2) - but only after checking the PORVs.
B. CORRECT. The question stem is developed from the description given in ND-93.3-LP-5, Surry Lesson Plan for Pressurizer Pressure Control, for a PT-445 failure with no operator actions. The lesson plan reads as follows:
- 7. P-445 fail high
- a. PORV 1456 will open causing pressure to decrease.
- b. When pressure reaches 2000 psig, the PORV will close.
- c. Pressure will stabilize around 2000 psig as the PORV opens and closes around the 2000 psig interlock.
With the PT-445 failure coincident with a main generator trip/reactor trip, PORV 1456 will open as pressure decreases (following Tave lowering/Lpzr lowering) but will close once pressure is less than 2000 psig. Therefore, when the RO is checking PZR Pressure during immediate operator actions of E-0 a Safety Injection will not be required, and the team will transition to 1-ES-0.1. As the PZR Pressure Control system responds to the transient, pressure will continue to INCREASE until pressure gets to approximately 2000 psig, and then it will oscillate around that value as PORV 1456 cycles open and closed. This is the condition as described in the current conditions; the RO happens to be monitoring pressure while the PORV is cycling open. The NEXT action in 1-ES-0.1 step 6. b) RNO 1) is that IF pressure is less than 2235 psig and decreasing, THEN do the following: 1) Verify PRZR PORVs closed. IF any valve can NOT be closed, THEN manually close associated block valve. There are then two acceptable (correct) procedural flow paths that could be used to mitigate the PT-445 failure. First, the SRO could directly enter 1-AP-31 to take actions for the failure.
Alternately, the SRO could enter 0-AP-53.00, which would then direct a transition to 1-AP-31 at step 7: GO TO THE APPROPRIATE ABNORMAL PROCEDURE: []
()-AP-31.00, Increasing or Decreasing RCS Pressure.
C. INCORRECT. As described above in the analysis of correct answer choice B.
For this distractor both parts (1) and (2) are incorrect, but plausible, because actions in 0-AP-53.00 would address both a failure of the spray valves and a failure of the PORV(s).
D. INCORRECT. As described above in the analysis of correct answer choice B.
For this distractor, the only incorrect statement is that A transition to 1-AP-31 will NOT be required. It is clear from the flow path through 0-AP-53.00 that AP-53.00 will take the initial actions to stabilize the plant, and then clearly directs the operator to transition to 1-AP-31. The transition statement (step 7) is not conditional: GO TO..
Supporting References
- 1. Surry Procedure 1-ES-0.1, REACTOR TRIP RESPONSE, rev. 44. Especially step 6.
- 2. Surry Procedure 0-AP-53.00, LOSS OF VITAL INSTRUMENTATION /
CONTROLS, rev. 14. The flow path would be steps 1-7.
- 4. Surry Lesson Plan ND-93.3-LP-5, PRESSURIZER PRESSURE CONTROL, rev.
13, dtd 12/04/08. Especially the description of PT-445 failure w/no operator action on page 12.
References Provided to Applicant None.
Answer: B
- 4. 000000000040AA2.04 1 Unit 1 Initial Conditions:
- Reactor power = 100%
- A steam line break develops inside containment Current Conditions:
- Reactor is tripped
- Containment pressure = 23 psia (maximum) and decreasing
- Pzr Level = 22% increasing E-2, FAULTED STEAM GENERATOR ISOLATION is in progress
- SCM=100oF Based on the above conditions, which one of the following states: (1) which SI initiation signal will be generated with 2 pressure switches at setpoint and (2) when transitioning from 1-E-2, to which procedure should the SRO transition?
A. (1) RCS Pressure Low (2) 1-ES-1.1, SI TERMINATION B. (1) RCS Pressure Low (2) 1-E-1, LOSS OF REACTOR OR SECONDARY COOLANT C. (1) Containment Pressure High (2) 1-ES-1.1, SI TERMINATION D. (1) Containment Pressure High (2) 1-E-1, LOSS OF REACTOR OR SECONDARY COOLANT K/A Steam Line Rupture - Excessive Heat Transfer: Ability to determine and interpret the following as they apply to the Steam Line rupture: Conditions requiring ESFAS initiation.
K/A Match Analysis To arrive at the correct answer, the applicant must interpret the excess cooling conditions, which requires SI, and select the appropriate procedure.
SRO Match Requires the applicant to know how to transition from one EOP section to another with varying plant conditions. Procedure selection beyond entry into major EOPs is being tested.
Answer Choice Analysis A. Incorrect: 1st part is correct. RCS pressure setpt is 2/3. 2nd part is incorrect because with degraded containment conditions, Pzr level is required to be [50%] before SI can be terminated.. Plausible because if degraded containment conditions were not present, it would be correct.
B. Correct: RCS pressure setpt is 2/3. The SRO will transition to 1-E-1 because with degraded containment conditions, Pzr level is required to be [50%] before SI can be
terminated.
C. Incorrect: Containment pressure is 3/4 pressure switches at setpt. Plausible because it is an initiation signal for SI. 2nd part is incorrect because with degraded containment conditions, Pzr level is required to be [50%] before SI can be terminated..
Plausible because if degraded containment conditions were not present, it would be correct.
D. Incorrect: Containment pressure is 3/4 pressure switches at setpt. Plausible because it is an initiation signal for SI. 2nd part is correct.
Supporting References 1-E-0 ND-95-3 Critical Safety Function Status Trees Obj: E ND-91-LP-2 Obj 1 ND-91-LP-3 ND-95.3-LP-46 FR-P.1 References Provided to Applicant none Answer: B
- 5. 000000000055EA2.03 1 Initial Conditions:
- Unit 1 is experiencing a sustained Loss of All AC Power condition.
- The TDAFW pump shaft sheared on startup, and all efforts to cross-connect AFW with Unit 2 have failed.
Current Conditions:
- All emergency buses remain de-energized.
- Operators have just completed the step in ECA-0.0 to Check DC Bus Loads,and have placed both the DC emergency oil pump and the Air Side seal oil backup pump in PTL.
- Core Exit Thermocouples (CETCs) are 1202 °F and rising.
- STA reports the following Critical Safety Function Status Trees:
- Core Cooling: RED
- Heat Sink: RED
- Containment: ORANGE
- Inventory: YELLOW
- Subcriticality: GREEN
- Integrity: GREEN
Based upon the current conditions, what procedure is required to be used NEXT to mitigate the casualty?
A. 1-FR-C.1, RESPONSE TO INADEQUATE CORE COOLING.
B. 1-ECA-0.0, LOSS OF ALL AC POWER.
C. 1-FR-H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK.
D. 1-SACRG-1, SEVERE ACCIDENT CONTROL ROOM GUIDELINE INITIAL RESPONSE.
K/A Station Blackout 055EA2.03 Ability to determine or interpret the following as they apply to a Station Blackout: Actions necessary to restore power.
(CFR 43.5 / 45.13) (SRO - 4.7)
K/A Match Analysis This question matches the K/A statement by requiring the SRO applicant to correctly interpret a set of given plant conditions during a Station Blackout (Loss of All AC Power) casualty, and then determine the appropriate procedure to take actions necessary to mitigate the accident. To arrive at the correct answer, the SRO applicant must recognize that in an ECA-0.0 scenario FR-P procedures are not implemented due to the inadequate electrical power situation, and also to recognize that conditions are met that require transitioning from ECA-0.0 to the SACRG procedures.
SRO Guidance Analysis The question author intends to link this question to 10CFR55.43(b)(5) - assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. In accordance with rev. 1 of the SRO-Only Guidance, Figure 2, the key assessment for this question is that arriving at the correctly answer requires detailed knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures.
A summary of the screening process is as follows:
- Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, location, etc.
Answer: NO
- Can the question be answered solely by knowing immediate operator actions?
Answer: NO
- Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?
Answer: NO
- Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?
Answer: NO
- Does the question require one or more of the following? [] Detailed knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures?
Answer: YES - may be considered SRO-only.
Answer Choice Analysis A. INCORRECT. The note at the beginning of ECA-0.0 states that FRP procedures should not be performed during a Loss of All AC Power casualty. However, this distractor is plausible because CETCS at 1200 F is a clear RED path transition to FR-C.1 per the CSFSTs.
B. INCORRECT. The question stem places the operator at the step in ECA-0.0 where a check is made to transition from the EOPs to the SACRGs. This answer is plausible if the applicant remembers that FRPs are not performed in ECA-0.0, but does not remember the correct transition criteria to SACRG-1.
C. INCORRECT. As per the note before step 1 of ECA-0.0, FR procedures are not to be implemented during ECA-0.0/Loss of All AC Power casualties. Also, the RED path in CORE COOLING (FR-C.1) would take precedence in any case over HEAT SINK.
However, this distractor is plausible because the question stem takes such pains to describe the AFW situation in the "initial conditions" section.
D. CORRECT. Step 29 RNO of ECA-0.0 requires a transition to SACRG-1 if CETCs are 1200 F and increasing.
Supporting References
- 1. Surry Procedures for Critical Safety Function Status Trees (F-1 through F-6).
- 3. Surry Procedure 1-ECA-0.0, "LOSS OF ALL AC POWER," rev. 30. Especially steps 28-29 on p. 21 of 24.
- 4. This question is modified from Surry 2004-301 055EA2.03 and Turkey Point 2005-301 055EA2.03 to change answer to SACRGs and to enhance plausibility.
References Provided to Applicant None.
Answer: D
- 6. 000000000WE011EA2.1 1 Unit 1 initial conditions:
- A large break LOCA occurred.
- The crew transitioned to ES-1.3, Transfer to Cold Leg Recirculation.
Current conditions:
- RWST level is 12% and dropping.
- Recirc Mode Transfer (RMT) switches have been placed in the RMT position.
- Amps and flow indications are oscillating on A and B LHSI pumps:
- Containment sump level is 54 inches.
- Crew has entered 1-ES-1.3, Attachment 1, Containment Sump Screen Blockage -
Contingency Actions.
Based on current conditions, which one of the following describes the actions as required by Attachment 1 of 1-ES-1.3?
A. Trip both LHSI pumps, cross-tie Unit 1 charging system with Unit 2 RWST and return to the procedure step in effect of 1-ES-1.3.
B. Trip one LHSI pump, cross-tie LHSI pump suction with Unit 2 RWST and return to the procedure step in effect of 1-ES-1.3.
C. Trip both LHSI pumps, cross-tie Unit 1 charging system with Unit 2 RWST and go to ECA-1.1, Loss of Emergency Coolant Recirculation.
D. Trip one LHSI pump, cross-tie LHSI pump suction with Unit 2 RWST and go to ECA-1.1, Loss of Emergency Coolant Recirculation.
K/A Loss of Emergency Coolant Recirculation
Ability to determine and interpret the following as they apply to the (Loss of Emergency Coolant Recirculation): Facility conditions and selection of appropriate procedures during abnormal and emergency operations.
(CFR: 43.5/45.13) (SRO - 4.2)
K/A Match Analysis The applicant must interpret the facility conditions (sump blockage on both LHSI pumps), determine the actions required by Attachment 1 of 1-ES-1.3 which will require selecting between transitioning to 1-ECA-1.1, Loss of Emergency Coolant Recirculation or returning to the main body of 1-ES-1.3.
SRO Guidance Analysis The question requires detailed knowledge of procedure steps and decision points within the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures. This is therefore testing procedural knowledge on a different and more detailed level than what is expected for a RO.
Answer Choice Analysis A. INCORRECT. Plausible because tripping both LHSI pumps and cross-tying the charging system to Unit 2 RWST are both actions directed by Attachment 1. Returning to the step in effect is an action within Attachment 1, however this would be incorrect with two LHSI pumps that have lost their suction source.
B. INCORRECT. Plausible because the first action is to trip one of the two LHSI pumps followed by cross-tying with Unit 2 RWST. Cross-tying to the LHSI pump suction is plausible because this would provide a source of water to inject through the LHSI piping via the cold legs. However the cross-tying involves the charging system rather than the LHSI system. Returning to the step in effect is an action within Attachment 1, however this would be incorrect with two LHSI pumps that have lost their suction source.
C. CORRECT. Attachment 1 performs the following sequence: trips one affected LHSI pump, crossties the charging system with Unit 2 RWST and then trips the second affected pump and sends the user 1-ECA-1.1.
D. INCORRECT. Plausible because the first action is to trip one of the two LHSI pumps followed by cross-tying with Unit 2 RWST. Cross-tying to the LHSI pump suction is plausible because this would provide a source of water to inject through the LHSI piping via the cold legs. However the cross-tying involves the charging system rather
than the LHSI system. The last part of the response is correct because Attachment 1 sends the user to 1-ECA-1.1 Supporting References
- 1. Surry lesson plan ND-95-3-LP-10, "Transfer to Cold Leg Recirculation," rev. 11.,
Obj. C.
- 2. 1-ES-1.3, Transfer to Cold Leg Recirculation, Rev. 18.
- 4. Surry Lesson Plan ND-93.3-LP-5, PRESSURIZER PRESSURE CONTROL, rev.
13, dtd 12/04/08. Especially the description of PT-445 failure w/no operator action on page 12.
References Provided to Applicant None.
Answer: C
- 7. 000000024AG2.4.11 1 Unit 1 initial conditions:
- The reactor failed to trip after receiving a trip signal
- SRO transitioned from 1-E-0, REACTOR TRIP OR SAFETY INJECTION, to 1-FR-S.1, RESPONSE TO NUCLEAR POWER GENERATION/ATWS.
- Reactor Power = 23% decreasing Current plant conditions:
- Emergency Boration Initiated
- Reactor power = 3% decreasing
- Intermediate range channels indicate negative SUR
- Operators are verifying the reactor subcritical at the end of 1-FR-S.1 Based on the current plant conditions, which one of the following states (1) the emergency boration guidance in 1-FR-S.1, and (2) the next correct procedure transition as directed by 1-FR-S.1?
A. 1) Boration should continue. Obtaining adequate shutdown margin is not required prior to any procedure transition. 2) Return to 1-E-0.
B. 1) Boration should continue. Obtaining adequate shutdown margin is not required prior to any procedure transition.
- 2) Remain in 1-FR-S.1, and inititiate Attachment 1, VERIFYING APPLICABLE STEPS OF 1-E-0.
C. 1) Obtaining adequate shutdown margin is required prior to any procedure transition. 2) Return to 1-E-0.
D. 1) Obtaining adequate shutdown margin is required prior to any procedure transition. 2) Remain in 1-FR-S.1, and inititiate Attachment 1, VERIFYING APPLICABLE STEPS OF 1-E-0.
K/A Emergency Boration:G2.4.11 Knowledge of abnormal condition procedures.
K/A Match Analysis Knowledge of abnormal condition procedures (EOPs) emergency boration requirements is required to arrive at the correct answer.
SRO Match Requires the applicant to know how to transition from one EOP section to another with varying plant conditions. This transition knowledge is beyond basic AOP/EOP entry conditions.
Answer Choice Analysis A. Correct: FR-S.1 Caution at step 18 states Boration should be continued to obtain adequate shutdown margin during subsequent actions. At 3% power you do not have adequate SDM. When power < 3% and negative SUR on IR, FR-S.1 directs returning to the procedure and step in effect (E-0).
B. Incorrect: 1st part is correct. 2nd part is incorrect because SI has not occurred.
2nd part is plausible because if SI had occurred, it would be correct.
C. Incorrect: 1st part is plausible because emergency boration is part of 1FR-S.1. To transition out of FR-S.1, you need to be less that 5% AND either neg SUR on IRs or 1.77% SDM (adequate). 2nd part is correct.
D. Incorrect: 1st part is plausible because emergency boration is part of 1FR-S.1. To transition out of FR-S.1, you need to be less that 5% AND either neg SUR on IRs or 1.77% SDM (adequate). 2nd part is incorrect because SI has not occurred. 2nd part is plausible because if SI had occurred, it would be correct.
Supporting References 1FR-S.1 RESPONSE TO NUCLEAR POWER GENERATION/ATWS ND-95.3-LP-36 Obj: B TS 3.2-4 References Provided to Applicant none Answer: A
- 8. 000000033AG2.4.30 1 The following sequence of events occurred on Unit 1:
- Time = 1400. Reactor Power is 10-8 Amps and stable for taking critical data.
- Time = 1401. IR N-35 indication fails LOW.
- Time = 1402. Reactor Operator reports N-35 control power AND instrument power fuses are blown.
- Time = 1403. Reactor Power is 10-8 Amps and stable.
Based on the given sequence of events, what offsite (i.e. external to Surry) notifications are required in accordance with VPAP-2802, Notifications and Reports?
Consider that all required internal notifications (i.e. to operations, engineering, and station management, etc.) are, or will be, made.
(REFERENCE PROVIDED)
One hour or less Greater than one hour External Notifications External Notifications A. None required None required B. None required Notification(s) required C. Notification(s) required Notification(s) required D. Notification(s) required None required K/A
Loss of Intermediate Range NI 033AG2.4.30 Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.
(CFR 41.10 / 43.5 / 45.11) (SRO - 4.1)
K/A Match Analysis This question matches the K/A statement by requiring the SRO applicant to correctly interpret a set of given plant conditions during a Loss of Intermediate Range NI casualty, and then determine the appropriate notifications that are required based on the analysis. To arrive at the correct answer, the SRO applicant must recognize that the loss of control and instrument power to N-35 should have caused a reactor trip, but did not; and that therefore an EAL is required (which is a less than one hour notification). At some point, the reactor will be manually tripped, which will require a four-hour notification.
SRO Guidance Analysis The question author intends to link this question to 10CFR55.43(b)(1) - conditions and limitations in the facility license [including reporting requirements]. The SRO applicant is required to possess a higher level of knowledge regarding the notification/reporting requirements than the RO applicant. No linking flowchart is provided in Region II SRO-only guidance for questions linked to 10CFR55.43(b)(1).
Answer Choice Analysis A. INCORRECT. Because the reactor did not automatically trip when it should have, the station is in at least an ALERT. If the operators were not successful in manually tripping the reactor, the station would meet the conditions for a SITE AREA EMERGENCY. Either case would require a one hour or less notification. Based on the ALERT, a report to the State of Virginia in accordance with section 6.3.5.b would also be required, which is an 8-hour report. Therefore, C is the only correct answer.
This distractor is plausible if the applicant incorrectly believes that the reactor should not have tripped on the NI-35 failure, and that the only thing that happened was an instrument failure.
B. INCORRECT. See analysis of A above. This distractor is plausible if the applicant applies the logic for a reactor tripthe notifications for a normal reactor trip w/no SI at power are a 4-hour notification for the RPS actuation while critical and the AFW auto-start signal on lo-lo S/G levels. However, the candidate who chooses this answer does not realize that the reactor should have automatically tripped, but did not.
C. CORRECT. See analysis of A above; at a minimum, both one hour or less reports and greater than one hour reports are required to be made for these plant conditions.
D. INCORRECT. This distractor is plausible if the applicant wrongly believes that initial notifications exempt the plant from conducting follow-up notifications.
Supporting References
- 1. Surry Procedure VPAP-2802, Notifications and Reports, rev. 33.
References Provided to Applicant VPAP-2802, pages 70-91.
Answer: C
- 9. 000000037AA2.11 1 Unit 1 current conditions:
- A steam generator (SG) tube leak exists on A SG.
- The reactor is tripped.
- SI is not required.
- The crew has initiated 1-AP-24.01, Large Steam Generator Tube Leak.
Which one of the following completes the statements below concerning the tube leak on A SG?
In accordance with 1-AP-24.01, SG level is required to be greater than (1) before stopping feed to the leaking SG AND Tave is required to be less than (2) before closing the associated main steam trip valve on the leaking SG.
A. (1) 12% Narrow Range (2) 543°F B. (1) 12% Narrow Range (2) 547°F C. (1) 22% Narrow Range (2) 543°F D. (1) 22% Narrow Range (2) 547°F K/A
Steam Generator Tube Leak Ability to determine and interpret the following as they apply to the Steam Generator Tube Leak: When to isolate one or more steam generators.
(CFR: 43.5/45.13) (SRO - 3.8)
K/A Match Analysis The SRO applicant must know the conditions that must be met before a leaking steam generator can be isolated.
SRO Guidance Analysis The question requires knowledge of specific conditions contained within AOP action steps. While the conditions are similar to the conditions contained in the EOP that addresses a ruptured steam generator (1-E-3, Ruptured Steam Generator), 1-AP-24.01 contains an additional requirement concerning a maximum allowed RCS Tave before the affected SGs steam line can be isolated. This is therefore testing procedural knowledge on a different and more detailed level than what is expected for a RO.
Answer Choice Analysis A. INCORRECT. Plausible because 12% narrow range (NR) level indication is correct. However, a Tave of 543°F is actually below the temperature requirement of Tave less than or equal to 547°F. Plausible because a Tave of 543°F is the temperature used by both 1-AP-24.01 and 1-E-3 to determine when to block the low Tave safety injection (SI) signal.
B. CORRECT. In accordance with step 9 of 1-AP-24.01 the affected SG must have minimum level of 12% on the NR level indication before stopping feed to the SG. Also, in accordance with step 18 of 1-AP-24.01 RCS Tave must be less than 547°F before closing the Main Steam Trip Valve (MSTV) on the affected SG.
C. INCORRECT. Plausible because 22% NR is lower limit of the desired level band for controlling feed flow on the intact SGs (step 20). In addition, a Tave of 543°F is the temperature used by both 1-AP-24.01 and 1-E-3 to determine when to block the low Tave safety injection (SI) signal.
D. INCORRECT. Plausible because 22% NR is lower limit of the desired level band for controlling feed flow on the intact SGs (step 20). The second part of the response is correct.
Supporting References
- 1. Surry lesson plan ND-95.2-LP-6, "Steam Generator Tube Rupture," rev. 29, OBJ. D, pp. 25-29.
- 2. 1-AP-24.01, Major Steam Generator Tube Rupture, Rev. 29, pp. 5-8.
References Provided to Applicant None.
Answer: B
- 10. 000000WE03EA2.2 1 Unit 1 conditions:
- A LOCA has occurred
- The crew is performing Step 4 of 1-ES-1.1, SI TERMINATION, which requires the operators to stop all but one charging pump and place them in AUTO.
- With only one charging pump now running, RCS pressure begins to decrease.
Based on the above conditions, which one of the following states the required action (if any) and the correct procedure implementation?
A. Manually reinitiate SI and transition to 1-E-0, REACTOR TRIP OR SAFETY INJECTION.
B. Manually restart a charging pump and monitor RCS pressure while continuing in 1-ES-1.1.
C. Transition to 1-ES-1.2, POST LOCA COOLDOWN AND DEPRESSURIZATION.
D. Continue in 1-ES-1.1. Restart of a charging pump is not required.
K/A LOCA Cooldown - Depress. / Ability to determine and interpret the following as they apply to the (LOCA Cooldown and Depressurization): Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.
K/A Match Analysis Question requires knowledge of adherence to EOPs for LOCA cooldown and
depressurization. The EOPs are required by Tech Specs, which is an appendix to the license; therefore, operation within the limits of the license is being tested.
SRO Guidance Analysis The question requires assessing plant conditions (normal, abnormal, or emergency) and then prescribing a procedure or section of a procedure to mitigate, recover, or with which to proceed. This test procedure entry beyond that of major EOPs.
Answer Choice Analysis A. INCORRECT: If RCS pressure cannot be maintained here then the RCS needs to be further cooled down and depressurized so the transition is made to ES-1.2, Post LOCA Cooldown and Depressurization.
B. INCORRECT. If RCS pressure cannot be maintained here then the strategy is to further cooldown and depressurized not start additional pumps or initiate SI.
C. CORRECT. With one charging pump running at this time in the event, if RCS pressure cannot be maintained further cooldown and depressurization is warranted so the transition is made to ES-1.2, Post LOCA Cooldown and Depressurization.
D. INCORRECT. Lowering RCS pressure is not the expected response at this point in the transient so ES-1.1 should be exited to further cooldown and depressurize per ES-1.2, Post LOCA Cooldown and Depressurization.
Supporting References
Reference:
ES-1.1 Lesson Plan Objective: ND-95.3-LP-8A, obj. A References Provided to Applicant None.
Answer: C
- 11. 000004G2.1.23 1 Initial conditions:
- Both units are at 100% power.
- The crew entered 1-AP-8.00, Loss of Normal Charging Flow
- Gas binding is suspected on the Unit 1 charging pumps.
- Charging and letdown have been secured on Unit 1.
- Venting of the charging pumps was attempted, but was unsuccessful.
Current conditions:
- Pressurizer level is 18% and dropping.
Based on the current conditions and in accordance with 1-AP-8.00, which one of the following describes:
(1) the reactor trip requirements for both units, and (2) whether performance of Attachment 3, Charging Pump Cross-Connect, of 1-AP-8.00 is required?
A. (1) A reactor trip is required on Unit 1. A trip of Unit 2 reactor is not required.
(2) Performance of Attachment 3 is required.
B. (1) A reactor trip is required on Unit 1 and Unit 2.
(2) Performance of Attachment 3 is required.
C. (1) A reactor trip is required on Unit 1. A trip of Unit 2 reactor is not required.
(2) Performance of Attachment 3 is NOT required.
D. (1) A reactor trip is required on Unit 1 and Unit 2.
(2) Performance of Attachment 3 is NOT required.
K/A Chemical and Volume Control System Ability to perform specific system and integrated plant procedures during all modes of operation.
(CFR: 41.10/43.5/45.2/45.6) (SRO - 4.4)
K/A Match Analysis The applicant must perform specific actions from 1-AP-8.00 based on the inability to re-establish charging flow and pressurizer level below 20%.
SRO Guidance Analysis The question requires knowledge of the specific content of the procedure and the attachments and based on the plant conditions determine whether procedural attachments need to be implemented.
Answer Choice Analysis A. INCORRECT. Plausible because a trip of Unit 1 is required, but under current plant conditions (i.e. pressurizer level less than 20%), both units are required to be tripped.
Also, performance of Attachment 3 is directed by 1-AP-8.00.
B. CORRECT. Steps 3 and 4 of 1-AP-8.00 direct performance of 1-E-0, 2-E-0 and . In addition, the procedural note prior to step 3 of 1-AP-8.00 states If in the judgment of the Shift Manager charging flow cannot be restored prior to PRZR level decreasing below 20%, both units must be tripped and Charging cross-connected IAW .
C. INCORRECT. Plausible because a trip of Unit 1 is required, but under current plant conditions both units are required to be tripped. The second half of the response is plausible if the applicant is not aware of the procedural note prior to step 3 requiring cross-connect of the units charging pumps.
D. INCORRECT. Plausible because tripping of both units is required. The second half of the response is plausible if the applicant is not aware of the procedural note prior to step 3 requiring cross-connect of the units charging pumps.
Supporting References
- 1. Surry lesson plan ND-88.3-LP-5, "Charging Pumps," rev. 17, Obj. E, pg. 13.
- 2. 1-AP-8.00, Loss of Normal Charging Flow, Rev. 12, pg. 4.
References Provided to Applicant None.
Answer: B
- 12. 000006A2.13 1 Unit 1 initial conditions:
- The Residual Heat Removal system is in service.
- At 10:03 A.M., a spurious Safety Injection (SI) actuation occurred on both trains of SI.
- The crew entered 1-AP-10.20, Response to Spurious Safety Injection with RCS Temperature Less Than 350°F.
Current conditions (10:05 A.M.):
- The SI Reset pushbuttons have been depressed on both trains.
- 1A-F3, SI INITIATED TRAIN A, is lit.
- A field operator has been sent to open the DC breaker for SI Train A.
Based on the current conditions, which one of the following identifies (1) the charging pump(s) feeding the reactor IMMEDIATELY after the SI initiation, and (2) the required procedural transition in accordance with 1-AP-10.20?
A. (1) One charging pump.
(2) Exit 1-AP-10.20 and Go to 1-AP-10.19, Resetting Safety Injection.
B. (1) Three charging pumps.
(2) Exit 1-AP-10.20 and Go to 1-AP-10.19, Resetting Safety Injection.
C. (1) One charging pump.
(2) Initiate 1-AP-10.19, Resetting Safety Injection and continue performing 1-AP-10.20.
D. (1) Three charging pumps.
(2) Initiate 1-AP-10.19, Resetting Safety Injection and continue performing 1-AP-10.20.
K/A Emergency Core Cooling Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations: Inadvertent SIS actuation.
(CFR: 41.5/45.5) (SRO - 4.2)
K/A Match Analysis The applicant must interpret the plant conditions (Intermediate Shutdown with reactor temperature less than 350°F and reactor pressure less than 450 psig when an inadvertent SI occurs that fails to reset on Train A) and predict the sources injecting to the reactor following an inadvertent SI. Based on the response of the SI system, the applicant must also determine the correct procedural transition to mitigate both the inadvertent SI and the failure of train A to reset.
SRO Guidance Analysis The question requires detailed knowledge of the content contained in both the unit shutdown procedures and abnormal operating procedures to recognize that at the current reactor pressure and temperature (i.e., less than 350°F / 450 psig) all but one charging pump will be in PTL as well as the required course of action associated with
RNO step 3 of 1-AP-10.20 as a result of one of the SI trains failing to reset. The level of detail being tested is beyond the knowledge of the mitigative strategy of the abnormal procedure or the purpose of the unit shutdown procedure and therefore appropriate as SRO-only testing concepts. Procedure selection beyond basic AOP entry conditions.
Answer Choice Analysis A. INCORRECT. Plausible because the first half of the response is correct (See discussion below for choice B The second half of the response is plausible since SI Train A did not reset and it would appear reasonable to exit the procedure mitigating the inadvertent SI (1-AP-10.20) until both trains of the SI logic system had been successfully reset.
B. INCORRECT. Plausible because the pump configuration would be correct if the inadvertent SI initiation had occurred with reactor temperature and pressure above 350°F and 450 psig. The second half of the response is plausible since SI Train A did not reset and it would appear reasonable to exit the procedure mitigating the inadvertent SI (1-AP-10.20) until both trains of the SI logic system had been successfully reset.
C. CORRECT. With reactor temperature less than 350°F and reactor pressure less than 450 psig, 2 of 3 Charging pumps are in PTL, so that on an inadvertent SI only one charging pump would be injecting into the reactor. In the event either or both trains of SI fails to reset then the RNO actions for step 3 of 1-AP-10.20 direct the initiation of 1-AP-10.19 and the user continues on with the subsequent actions of 1-AP-10.20 D. INCORRECT. Plausible because the pump configuration would be correct if the inadvertent SI initiation had occurred with reactor temperature and pressure above 350°F and 450 psig. The second half of the response is correc (See discussion above for choice c).
Supporting References
- 1. Surry lesson plan ND-91-LP-3, "Safety Injection System Operations," rev. 22, Obj. H, pp. 22-24.
- 2. 1-AP-10.20, Response to Spurious Safety Injection with RCS Temperature less than 350°F, Rev. 4, pg. 2.
- 3. 1-GOP-2.4, Unit Shutdown, HSD to 351°F., Rev. 41
- 4. 1-GOP-2.5, Unit Shutdown, 351°F to less than 205°F, Rev. 24, pg. 9 References Provided to Applicant None.
Answer: C
- 13. 000013A2.03 1 Unit 1 initial conditions:
- Reactor power = 100%
- RCS pressure = 1900 psig decreasing rapidly Current Conditions:
- Containment pressure = 22 psia (maximum) decreasing
- RCS pressure = 300 psig decreasing
- RWST level = 18% decreasing E-1 LOSS OF REACTOR OR SECONDARY COOLANT in progress Based on the above conditions, which one of the following states (1) the minimum operating containment spray pumps required to be operating as defined in FSAR Chapter 6, and (2) to what procedure 1-E-1 directs you to transition?
A. (1) One (2) 1-ES-1.2, POST LOCA COOLDOWN AND DEPRESSURIZATION B. (1) One (2) 1-ES-1.3, TRANSFER TO COLD LEG RECIRCULATION C. (1) Two (2) 1-ES-1.2, POST LOCA COOLDOWN AND DEPRESSURIZATION D. (1) Two (2) 1-ES-1.3, TRANSFER TO COLD LEG RECIRCULATION K/A Engineering Safety Features Actuation: Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS: and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Rapid Depressurization K/A Match Analysis Question requires knowledge of ES system operation during a rapid RCS depressurization
SRO Guidance Analysis Question requires knowledge of decision points in the EOP that involve transition to other EOP sections.
Answer Choice Analysis A. INCORRECT. 1st part is correct. 2nd part is incorrect because with adverse containment conditions [20 psia], the criteria for transitioning to ES-1.2 is [400] psig which is not met. 2nd part is plausible because if adverse containment conditions were not met or RCS pressure were > 400 psig, it would be correct.
B. CORRECT. UFSAR Ch 6 assumes a loss of one emergency bus which leaves 1 CS pump. With adverse containment conditions [20 psia], the criteria for transitioning to ES-1.2 is [400] psig which is not met. The next step in the procedure is if RWST is <
20% to transition to ES-1.3.
C. INCORRECT. UFSAR Ch 6 assumes a loss of one emergency bus which leaves 1 CS pump. 1st part is plausible because Ch 6 assumes 2 recirculation spray pumps.
2nd part is incorrect because with adverse containment conditions [20 psia], the criteria for transitioning to ES-1.2 is [400] psig which is not met. 2nd part is plausible because if adverse containment conditions were not met or RCS pressure were > 400 psig, it would be correct.
D. INCORRECT. UFSAR Ch 6 assumes a loss of one emergency bus which leaves 1 CS pump. 1st part is plausible because Ch 6 assumes 2 recirculation spray pumps.
2nd part is correct.
Supporting References ND-91-LP-2 SAFETY INJECTION pg 4 Obj: A ND-91-LP-3 SAFETY INJECTION OPERATION pg 4 1-E-1 LOSS OF REACTOR OR SECONDARY COOLANT References Provided to Applicant None.
Answer: B
- 14. 000064G2.1.20 1 Unit 1 initial conditions:
- A loss of offsite power has occurred.
- The crew has entered 1-E-0
- Emergency diesel generators (EDG) #1 and #3 have started and tied to their respective emergency buses.
Current conditions:
- Breaker 15J3, EDG #3 Emergency Supply breaker has tripped.
- The following alarms are lit on annunciator panel VSP-B5 - EMERG GEN 3 TRBL VSP-A5 - EMERG GEN 3 DIFF K-G4 - 4KV EMERG BUS EMERG SUP AUTO TRIP
- The exciter field breaker for EDG #3 is open.
- EDG #3 is in its cooldown cycle.
- The crew enters 0-AP-17.05, EDG 3 - Emergency Operations.
Based on these current conditions, which one of the following identifies (1) the FIRST attachment to be performed in accordance with 0-AP-17.05, and (2) the condition that tripped breaker 15J3?
A. (1) Attachment 1, Auxiliary Trip Relay Actuation - Contingency Actions.
(2) Field voltage was lost to EDG #3.
B. (1) Attachment 1, Auxiliary Trip Relay Actuation - Contingency Actions.
(2) EDG #3 differential lockout relay actuated.
C. (1) Attachment 3, Stripping the 1J Bus.
(2) Field voltage was lost to EDG #3.
D. (1) Attachment 3, Stripping the 1J Bus.
(2) EDG #3 differential lockout relay actuated.
K/A Emergency Diesel Generator Ability to interpret and execute procedure steps.
(CFR: 41.5/43.2/45.2) (SRO - 4.6)
K/A Match Analysis The SRO applicant must interpret the plant conditions and recognize that Attachment 1 must be performed first as a result of the actuation of the auxiliary trip relay. The applicant must also recognize that based on the plant conditions the differential lockout relay has actuated.
SRO Guidance Analysis The question focuses on selection and sequencing of procedural attachments and is therefore testing procedural knowledge on a different and more detailed level than what is expected for a RO.
Answer Choice Analysis A. INCORRECT. Plausible because the first half of the response is correct (See discussion below for choice B). Also, a loss of field voltage is plausible since it would cause an actuation of the auxiliary trip relay which in turn would trip breaker 15J3.
However, the EDG will not enter the cooldown cycle on a loss of field voltage.
B. CORRECT. Step 5 RNO directs performance of Attachment 1 if the auxiliary trip relay light is lit on EDG #3 control panel. In addition, actuation of the differential lockout relay will result in all of the alarms. Also, the EDG cooldown cycle is initiated anytime a 15J3 trips due to a differential lockout relay or an overcurrent trip.
C. INCORRECT. Plausible because Attachment 3 is required to be performed as a result of 1J Bus being de-energized (step 10 RNO). However, as a result of the actuation of the auxiliary trip relay, Attachment 1 will be performed as directed by Step 5 RNO. Also, a loss of field voltage is plausible since it would cause an actuation of the auxiliary trip relay which in turn would trip breaker 15J3, but the EDG will not enter the cooldown cycle on a loss of field voltage.
D. INCORRECT. Plausible because Attachment 3 is required to be performed as a result of 1J Bus being de-energized (step 10 RNO). However as a result of the actuation of the auxiliary trip relay, Attachment 1 will be performed as directed by Step 5 RNO.
The second half of the response is correct (See discussion above for choice B).
Supporting References
- 1. Surry lesson plan ND-90.3-LP-1, "Emergency Diesel Generators," rev. 17, Obj. E, pp.
27-29.
- 2. 0-AP-17.05, EDG 3 - Emergency Operations, Rev. 18, pg. 3.
- 3. 0-VSP-A5, Emerg Gen 3 Diff, Rev.1, pg. 2 of 3
Answer: B
- 15. 000078A2.01 1 Unit 1 Current Conditions:
- Reactor Power is 35%
- An Instrument Air Dryer malfunction occurs
- Operators have attempted, but could not bypass the dryer
- Instrument Air Pressure is 45 psig and stable
- Operators have just started taking actions in 1B-E6, IA LO HDR PRESS/IA COMPR 1 TBL Based on the current conditions, which one of the following:states the correct procedures to use to address the reduced Instrument Air Pressure, andstates whether the current conditions will result in a four hour notification being required in accordance with VPAP-2802, Notifications and Reports.
(REFERENCE PROVIDED)
A. (1) Go to AP-40.00, Non-recoverable Loss of Instrument Air, in conjunction with the EOP network of procedures.
(2) Four hour notification is required.
B. (1) Do NOT perform AP-40.00, Non-recoverable Loss of Instrument Air, in conjunction with the EOP network of procedures.
(2) Four hour notification is required.
C. (1) Perform AP-40.00, Non-recoverable Loss of Instrument Air, in conjunction with the EOP network of procedures.
(2) Four hour notification is NOT required.
D. (1) Do NOT perform AP-40.00, Non-recoverable Loss of Instrument Air, in conjunction with the EOP network of procedures.
(2) Four hour notification is NOT required.
K/A Instrument Air System (IAS)
Ability to (a) predict the impacts of the following malfunctions or operations on the IAS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Air Dryer and Filter Malfunctions.
K/A Match Analysis
The question requires knowledge of procedures to address the air dryer malfunction, which matches the second part of the K/A.
SRO Guidance Analysis The knowledge requirements of the question are linked to 10 CFR 55.43 via procedure selection. It would be an SRO job function to prioritize the use of an AP when the EOP network is in use. Also, the AP entry conditions pertinent to this question do not appear as basic entry criteria for this procedure - rather the entry to the AP is directed by the Alarm Response Procedure. Lastly, knowledge of notifications are typically the sole responsibility of the SROs on a crew.
Answer Choice Analysis A. CORRECT. 1B-E6 directs entry to AP-40, which then requires the reactor trip.
The four hour notification would be required by statement on page 82 of VPAP-2802.
B. INCORRECT. AP-40 not being performed is plausible because the EOP network has been entered, which is usually a higher priority than an AP. Secondly, the entry condition for AP-40 is located in the ARP. If the applicant only knows the entry conditions as listed in the AP, then the answer would not be known.
C. INCORRECT. The four hour notification not being required is plausible because the question clearly states that an automatic RPS actuation did not occur. The applicant must understand the requirement in the VPAP and correctly apply it to a manual reactor trip.
D. INCORRECT. See above.
Supporting References VPAP-2802, Notifications and Reports, Rev 33.
1B-E6, IA LO HDR PRESS/IA COMPR 1 TBL, Rev 12 References Provided to Applicant VPAP-2802: ONLY pages (1) Verify that use of AP-40 in conjunction with the EOPs is a correct answer. Also, help in identifying the supporting administrative requirements to support this answer would be appreciated.
(2) Verify that a manual trip is considered an RPS actuation.
Answer: A
- 16. 001A2.19 1 The following sequence of events occurs on Unit 1:
- Time = 1800. During a power increase, the unit stabilizes power at 85% to perform a calorimetric. The I target is 0.0%. I readings have remained in the target band for the past 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
- Time = 1801. A rod control circuit malfunction causes rods to insert.
- Time = 1802. Rod motion is stopped. I = -13.0%. Reactor Power = 80%.
- Time = 1817. I = -13.0%. Reactor Power = 85%.
- Time = 1837. I = -10.0%. Reactor Power = 85%.
- Time = 1905. I = -6.0%. Reactor Power = 85%.
- Time = 1910. I = -5.0%. Reactor Power = 85%.
Based on the given sequence of events, which one of the following answers (1) at Time
= 1911, is reactor power allowed to be raised above 90% in accordance with Technical Specification (TS) 3.12, CONTROL ROD ASSEMBLIES AND POWER DISTRIBUTION LIMITS, AND (2) the reason(s) for the above answer, in accordance with TS 3.12 BASIS?
(REFERENCE PROVIDED)
A. (1) No.
(2) Axial Xenon distribution control at less than 50% power is not as significant as axial Xenon control at full power, and allowances were made in the accident analyses (which is the basis of the I control procedures) for heat flux peaking factors for accidents occurring at less than 50% power.
B. (1) No.
(2) Axial Xenon distribution in the core has been affected to an extent that reactor power and reactor trip setpoint reductions are required.
C. (1) Yes.
(2) Axial Xenon distribution control at less than 90% power is not as significant as axial Xenon control at full power, and allowances were made in the accident analyses (which is the basis of the I control procedures) for heat flux peaking factors for accidents occurring at less than 90% power.
D. (1) Yes.
(2) Axial Xenon distribution in the core was NOT affected sufficiently to change the heat flux peaking factors which can be reached on a subsequent return to full power within the target band.
K/A 001 Control Rod Drive System (CRDS) 001A2.19 Ability to (a) predict the impacts of the following malfunction or operations on the CRDS- and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Axial flux distribution.
(CFR 41.5 / 43.5 / 45.3 / 45.13) (SRO - 4.0)
K/A Match Analysis This question matches the K/A statement by requiring the SRO applicant to apply Technical Specifications action statement to an operationally valid situation where a CRDS malfunction has caused axial flux distribution to go out of band. The SRO applicant must also apply knowledge of the TS basis to axial flux control to arrive at the correct reason for the TS actions (in a way, the knowledge of the TS basis is the prediction of the impacts of the CRDS malfunction).
SRO Guidance Analysis The question author intends to link this question to 10CFR55.43(b)(2), Facility operating limitations in the TS and their bases. Answers to the SRO clarification guidance questions are as follows:
Can question be answered solely by knowing = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action? Answer: NO.
Can question be answered solely by knowing the LCO/TRM information listed above-the-line? Answer: NO Can question be answered solely by knowing the TS Safety Limits? Answer: NO Does the question involve one or more of the following for TS, TRM, or ODCM?
- Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
- Knowledge of TS bases that is required to analyze TS required actions and terminology Answer: YES, question may be suitable for SRO-only question linked to 55.43(b)(2)
Answer Choice Analysis A. INCORRECT. Part (1) of this distractor is correct. The Control Rod malfunction caused delta-flux values to go outside the target band, but remain within the ACCEPTABLE OPERATION region of the TS Fig. 3.12-3. Therefore, the unit began accumulating penalty minutes as of either 1801 or 1802 on a one-for-one basis. At time 1901 or 1902, the unit is still outside the target band and has therefore greater than 60 penalty minutes in a 24-hour period, and is therefore in violation of TS 3.12.B.4.b.(1). The required action is to reduce power to less than 50%, and to reduce the high neutron flux setpoint to 55%. Part (2) of this distractor is incorrect. The key concept is stated in the TS BASIS to 3.12 as follows: The procedures for axial power distribution control are designed to minimize the effects of xenon redistribution on the axial power distribution during load-follow maneuvers. Therefore, the delta-flux procedures are not based on accident analyses concerns, they are based on load-follow maneuvers and the resulting Xenon transients. Part (2) is plausible because TS BASIS to 3.12 does state the following: xenon distribution control at part power is not as significant as the control at full power and allowance has been made in predicting the heat flux peaking factors for less strict control at part power.
Surry: please double-check that part (2) is incorrect enough. possible alternative: accident analyses concerned with target band, or only with
'doghouse'/acceptable operation band.
B. CORRECT. Parts (1) and (2) correct. Concerning Part (2), the TS BASIS states:
In some instances of rapid unit power reduction automatic rod motion will cause the flux difference to deviate from the target band when the reduced power level is reached.
This does not necessarily affect the xenon distribution sufficiently to change the envelope of peaking factors which can be reached on a subsequent return to full power within the target band. However, to simplify the specification, a limitation of one hour is any period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is placed on operation outside the band. This ensures that the resulting xenon distributions are not significantly different from those resulting from operation within the target band. Therefore, especially considering the last sentence, if the one-hour limitation is violated, the resulting xenon distributions have been affected to a point where power and high flux setpoints must be reduced, which is stated by the required action of the specification.
C. INCORRECT. Parts (1) and (2) are incorrect. Part (1) is plausible if the applicant mis-applies the TS 3.12, mis-calculates the accumulated penalty minutes, or incorrectly believes that since delta-flux values have returned to target the requirements of specification 3.12.B.4.(b) no longer applies. Further plausible because Part (1) is exactly the words of TS 3.12.B.4.c. Part (2) is plausible, a re-wording of the second part of distractor A. See analysis of A part (2) above.
D. INCORRECT. Parts (1) and (2) are incorrect. Part (1) is the same as C. above.
Part (2) would be correct if part (1) was correct, and is therefore plausible; the words are taken directly from the TS 3.12 BASIS.
Supporting References
- 1. Surry Technical Specification (TS) 3.12, CONTROL ROD ASSEMBLIES AND POWER DISTRIBUTION LIMITS, and the associated TS BASIS. Amendments Nos.
265 and 264.
- 2. Surry Lesson Plan ND-93.2-LP-4, POWER RANGE NIS, rev. 21. Especially p. 29 discussion on delta flux TS.
References Provided to Applicant Surry TS 3.12 pages 3.12-5, 3.12-6, and TS Figure 3.12-3.
Answer: B
- 17. 017G2.4.18 1 Unit 1 initial conditions:
- The reactor was tripped due to a tube rupture on B steam generator (SG).
- A Safety Injection (SI) initiated. E-3, Steam Generator Tube Rupture, is being performed.
- RCS cooldown to target temperature of 495°F has been reached.
Current conditions:
- B SG pressure is 1020 psig and stable.
- SI has been reset.
- LHSI pumps have been reset.
- RCS subcooling is 35°F.
- Operators are evaluating RCS subcooling in accordance with 1-E-3 before continuing with RCS depressurization.
Based on these current conditions, which one of the following describes (1) whether a procedure transition to 1-ECA-3.1, SGTR with Loss of Reactor Coolant - Subcooled Recovery is required AND (2) the reason for the actions based on the subcooling evaluation in accordance with 1-E-3?
A. (1) Transition to 1-ECA-3.1 is required.
(2) Subcooling cannot be assured and actions must be taken to re-establish subcooling.
B. (1) Transition to 1-ECA-3.1 is NOT required.
(2) Subcooling cannot be assured and actions must be taken to re-establish subcooling.
C. (1) Transition to 1-ECA-3.1 is required.
(2) Loss of RCS coolant from other than the tube rupture may be occurring.
D. (1) Transition to 1-ECA-3.1 is NOT required.
(2) Loss of RCS coolant from other than the tube rupture may be occurring.
K/A In-core Temperature Monitor Knowledge of the specific bases for EOPs.
(CFR: 41.10/43.1/45.13) (SRO - 4.0)
K/A Match Analysis The applicant must recognize that a subcooling margin of 50°F has not been met per EOP 1-E-3, Steam Generator Tube Rupture and determine that transition to 1-ECA-3.1 is required and the basis of this action in the EOP.
SRO Guidance Analysis This question requires the applicant to recognize the general location within 1-E-3 (after initial RCS cooldown and before RCS depressurization continues), apply the plant conditions and recognize that transition to an emergency contingency procedure (1-ECA-3.1) is required. The question requires detailed knowledge of the procedural steps and when procedural transitions are required. This is beyond the overall mitigative strategy of the EOP and would not be knowledge expected of an RO applicant. In addition, the SRO applicant must also know the basis for taking action.
Answer Choice Analysis A. INCORRECT. Plausible because the procedure transition portion of the response is correct. The second half of the response is plausible because this explanation is consistent with the reasons used throughout the EOPs for ensuring a subcooling margin above 30°F.
B. INCORRECT. Plausible because the subcooling margin is above the subcooling limit used in other steps of the procedure (i.e., 30°F) and would indicate that the subcooling margin is adequate to continue in 1-E-3. However, the subcooling limit after the initial RCS cooldown is 50°F. The second portion of the response is plausible because this explanation is consistent with the reasons used throughout the EOPs for ensuring a subcooling margin above 30°F.
C. CORRECT. Once the target temperature has been reached and RCS cooldown has been stopped, Step 18 of 1-E-3 directs the user to check if RCS subcooling is
greater than 50°F. The RNO for this step directs the user to go to 1-ECA-3.1. The basis for the action is that once RCS Cooldown is complete a subcooling margin of less than 50°F indicates that a loss RCS inventory is due to more than a just the ruptured steam generator.
D. INCORRECT. Plausible because the subcooling margin is above the subcooling limit used in other steps of the procedure (i.e., 30°F) and would indicate that the subcooling margin is adequate to continue in 1-E-3. However, the subcooling limit after the initial RCS cooldown is 50°F. The second half of the response is correct (See discussion in choice C above.).
Supporting References
- 1. Surry lesson plan ND-95.3-LP-13, "1-E Steam Generator Tube Rupture," rev. 17, Objectives B & C, pgs. 34 and 35.
- 2. 1-E-3, Steam Generator Tube Rupture, Rev. 40, pg. 14.
References Provided to Applicant None.
Answer: C
- 18. 086A2.01 1 Plant Conditions:
- Plant fire protection systems were manually disabled to perform unplanned maintenance.
- The disabled fire protection systems allowed a small fire to escalate into a large fire in the MCR.
- The CRS determines that MCR must be evacuated.
- The Auxiliary Shutdown Panel is available.
Which one of the following correctly (1) describes the entry into 0-FCA-1.00, Limiting MCR Fire, and (2) states whether EOP actions or FCA actions take precedence when conflicting guidance is encountered?
A. (1) Directly enter 0-FCA-1.00 prior to going to 0-AP-48.00, Fire Protection -
Operations Response.
(2) EOP actions take precedence over FCA actions when conflicting guidance is encountered.
B. (1) Directly enter 0-FCA-1.00 prior to going to 0-AP-48.00, Fire Protection -
Operations Response.
(2) FCA actions take precedence over EOP actions when conflicting guidance is encountered.
C. (1) First enter 0-AP-48.00, then enter 0-FCA-1.00, Fire Protection - Operations Response.
(2) EOP actions take precedence over FCA actions when conflicting guidance is encountered.
D. (1) First enter 0-AP-48.00, then enter 0-FCA-1.00, Fire Protection - Operations Response.
(2) FCA actions take precedence over EOP actions when conflicting guidance is encountered.
K/A Fire Protection System 086A2.01 Ability to (a) predict the impacts of the following malfunctions or operations on the Fire Protection System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Manual shutdown of the FPS.
K/A Match Analysis The question requires knowledge of how to use procedures, namely FCA and EOPs, when the FPS is shutdown and as a result a fire causes the control room to be evacuated.
SRO Guidance Analysis Knowledge of procedure selection is required. The required knowledge goes beyond simply AOP entry conditions or entry to major EOPs. The applicant must have knowledge of the internal procedure transition from AOP-48 to FCA-1.00. The question also tests procedure priority when conflicting EOP and FCA guidance exists.
Answer Choice Analysis A. Incorrect. Both parts are incorrect as described in "D" below. The first part is plausible because the fire is in the MCR and it would be logical to place limiting the fire as a high priority. The second part is plausible because typically the EOP network takes priority over all other procedure categories.
B. Incorrect. See above.
C. Incorrect. See Above.
D. Correct. Entry to FCA-1.00 if directed by AP-48.00, there is no direct entry into FCA-1.00. Also, a note in AP-48.00 states that when conflicting guidance is encounterd, the FCA takes precedents.
Supporting References
- 1. 0-AP-48.00, Fire Protection - Operations Response, Rev. 23.
- 2. 0-FCA-1.00, Limiting MCR Fire, Rev. 42.
References Provided to Applicant None.
Answer: D
- 19. G2.1.1 1 Initial Conditions:
- Unit 1 is in a scheduled refueling outage, with Unit 1 in the REFUELING SHUTDOWN mode as defined by Technical Specifications (TS).
- Unit 2 is in an unscheduled outage, with Unit 2 in the COLD SHUTDOWN mode, to perform emergent repairs on a leaking Pressurizer safety valve.
- The Shift Manager (SM) is filling the Command Function in the control room.
Current Conditions:
- The SM receives a phone call from the Plant Manager directing him to attend an outage risk assessment meeting in ten minutes. The meeting will be held in the admin building [Surry input correct nomenclature here].
Based on the given conditions, which one of the following is the required delegation of authority in accordance with TS?
A. The SM is required to designate an individual with a valid Senior Reactor Operator (SRO) license to assume the control room Command Function. An individual with a valid Reactor Operator license is NOT allowed to assume the Command Function.
The Shift Technical Advisor is NOT allowed to assume the Command Function.
B. The SM is required to designate an individual with either a valid Reactor Operator license, or a valid Senior Reactor Operator license, to assume the control room Command Function. The Shift Technical Advisor is NOT allowed to assume the Command Function.
C. The SM is required to designate an individual with either a valid Senior Reactor Operator license, or the Shift Technical Advisor, to assume the control room Command Function. An individual with a valid Reactor Operator license is NOT allowed to assume the Command Function.
D. The SM is required to designate an individual with either a valid Reactor Operator license, or a valid Senior Reactor Operator license, or the Shift Technical Advisor to assume the control room Command Function.
K/A Generic: Conduct of Operations G2.1.1 Knowledge of conduct of operations requirements.
(CFR 41.10 / 45.13) (SRO - 4.2)
K/A Match Analysis This question matches the K/A statement by requiring the SRO applicant to correctly determine the correct delegation of authority of the control room Command function given a set of operationally valid plant conditions. To arrive at the correct answer, the SRO applicant must understand the basis (in this case, the explanatory notes) of Table 6.1-1, MINIMUM SHIFT CREW COMPOSITION, which specifies that During any absence of the Shift Manager from the Control Room while the unit is shutdown or refueling, an individual with a valid SRO or RO license (other than the technical advisor) shall be designated to assume the Control Room command function.
SRO Guidance Analysis The question author intends to link this question to 10CFR55.43(b)(1) - conditions and limitations in the facility license. The SRO applicant is required to possess a higher level of knowledge regarding the Technical Specifications requirements for shift staffing, especially in an off-normal conditions such as is provided in the question. Although no linking flowchart is provided in the SRO-only clarification guidance for questions linked to 10CFR55.43(b)(1), part II. A. fourth bullet states: Some examples of SRO exam items for this topic include: The required actions for not meeting administrative controls listed in Technical Specification (TS) section 5 or 6, depending on the facility (e.g., shift staffing requirements). Therefore, the questions meets SRO-only guidance requirements.
Answer Choice Analysis A. INCORRECT. Although the first sentence in this distractor is correct, in that the SM may designate an SRO to hold the command function in his absence, the second sentence is incorrect, because an RO may also hold the command function in the given plant condition (two units in cold shutdown or refueling). The explanatory notes of Table 6.1-1, MINIMUM SHIFT CREW COMPOSITION, in TS states the following:
During any absence of the Shift Manager from the Control Room while the unit is shutdown or refueling, an individual with a valid SRO or RO license (other than the technical advisor) shall be designated to assume the Control Room command functions. This distractor is plausible because it is correct for most normal plant conditions; i.e. at least one unit operating.
B. CORRECT. See analysis of A above.
C. INCORRECT. See analysis of A above. The technical advisor is not allowed to assume the command function. This distractor is plausible if the applicant mis-understands the requirements for two units shutdown/refueled, or if an applicant believes that because a dedicated STA is not required under these conditions, the SM may delegate the command function to a non-required STA.
D. INCORRECT. See analysis of A above. The SM must delegate the command function to either an RO or an SRO. This distractor is plausible because there are several shift positions, including the SRO position and the STA position, that are not required to be filled under the current plant conditions; and it is plausible to wrongly assume that because the SRO position does not have to be filled, then the control room command function may not have to be filled.
Supporting References
- 1. Dominion Administrative Procedure OP-AA-100, Conduct of Operations, rev. 7.
Especially attachment 2 p. 36 of 72.
- 2. Surry Technical Specifications Table 6.1-1, MINIMUM SHIFT CREW COMPOSITION, p. TS 6.1-3, in its entirety.
References Provided to Applicant None.
Answer: B
- 20. G2.1.42 1 In accordance with 1-OP-FH-001, Controlling Procedure for Refueling, which one of the following identifies when subcriticality multiplication monitoring must be initiated during core reload?
A. Once four assemblies have been loaded in the core.
B. Once six assemblies have been loaded in the core.
C. Once eight assemblies have been loaded in the core.
D. Once nine assemblies have been loaded in the core.
K/A Plant-wide Generics Knowledge of new and spent fuel movement procedures.
(CFR: 41.10/43.5/45.12) (SRO - 3.4)
K/A Match Analysis The applicant must recognize when the procedural requirement for initiating subcriticality multiplication during core reload operations is required.
SRO Guidance Analysis The applicant must recognize whether surveillance requirements in support of core reload, specifically subcriticality multiplication monitoring are conducted at the appropriate time in the refueling process. Aministratively, the SRO is required to know when the monitoring is required to begin.
Answer Choice Analysis A. INCORRECT. Plausible because chi-squared test for SRNI-N-31 is required once four assemblies have been loaded per step 5.4.57.a of 1-OP-FH-001.
B. INCORRECT. Plausible because chi-squared test for SRNI-N-32 is required once six assemblies have been loaded per step 5.4.57.b of 1-OP-FH-001.
C. INCORRECT. Plausible because a Caution statement prior to step 5.4.55 of 1-OP-FH-001 that the chi-squared test may be done before the eighth fuel assembly is loaded in the core and must be done within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of loading the ninth assembly.
D. CORRECT. Per step 5.4.61, subcritical multiplication monitoring must start once the ninth assembly is on-loaded.
Supporting References
- 1. Surry lesson plan ND-92.5-LP-1, "Refueling Overview," rev. 18, Objectives B & E, pg. 11.
- 2. 1-OP-FH-001, Controlling Procedure for Refueling, Rev. 23, pp. 62-64.
References Provided to Applicant None Answer: D
- 21. G2.2.18 1 Initial Conditions:
- Unit 2 is in a refueling outage.
Current Conditions:
- The Shift Technical Advisor (STA) identifies that, due to multiple schedule changes and emergent switchyard work, there is a two-hour window that will occur several hours later on the current shift that would place the unit in a RED Shutdown Risk (SDR) condition.
Based on the current conditions, which one of the following are the MINIMUM required actions before entry into the RED SDR condition is allowed (if any), in accordance with OU-AA-200, Shutdown Risk Management?
The STA is required to inform the Shift Manager and Shift Outage Manager/Coordinator, and ensure ________________________________ .
A. they declare these associated work activities are High Risk Evolutions. Implement a SDR Contingency Plan approved by the Plant Manager and Facility Safety Review Committee. Establish Protected Trains/Equipment for the affected Key Safety Function(s). Entry into the RED SDR condition is only allowable if the planned duration is less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, AND the unit is NOT in a mid-loop condition with fuel in the reactor vessel.
B. they declare these associated work activities are High Risk Evolutions. Implement a SDR Contingency Plan approved by the Plant Manager and Facility Safety Review Committee. Establish Protected Trains/Equipment for the affected Key Safety Function(s). Once these actions are taken, OU-AA-200 does not specify any time restrictions or unit configuration restrictions once the RED SDR condition is entered.
C. they declare these associated work activities are High Risk Evolutions. Ensure a SDR Contingency Plan for the given conditions is approved by the Plant Manager and Facility Safety Review Committee, but implementation of the Contingency Plan is NOT required prior to entry into the RED SDR condition. Establish Protected Trains/Equipment for the affected Key Safety Function(s). Entry into the RED SDR condition is only allowable if the planned duration is less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, AND the unit is NOT in a mid-loop condition with fuel in the reactor vessel.
D. they understand that entry into the RED SDR condition is NOT allowed, even under an approved SDR Contingency Plan and with several layers of compensatory measures in place. The activities that would place the station in this configuration must be rescheduled.
K/A Generic: Equipment Control G2.2.18 Knowledge of the process for managing maintenance activities during shutdown operations, such as risk assessments, work prioritization, etc.
(CFR 41.10 / 43.5 / 45.13) (SRO - 3.9)
K/A Match Analysis This question matches the K/A statement by requiring the SRO applicant to correctly determine the requirements for a planned entry into a RED Shutdown Risk (SDR) condition during a refueling outage. The SRO applicant needs to recognize that the conditions in the question constitute a planned entry into a RED SDR, and that by procedure, a planned entry into a RED SDR is not allowed under any circumstances.
The distractors use elements that are correct for an unplanned entry into an ORANGE SDR condition for plausibility.
SRO Guidance Analysis The question author intends to link this question to 10CFR55.43(b)(1) - conditions and limitations in the facility license. As a note, the SRO clarification guidance document does not include a flow chart for this topic. In effect, the limitations listed in the Dominion procedure OU-AA-200 add additional limitations upon the station above and beyond the strict license basis. Furthermore, it is clear that the SRO-level individual is the expected user of this procedure (and therefore knowledge of its administrative requirements is expected at the SRO-only level, and NOT at the RO- or systems-level).
Answer Choice Analysis A. INCORRECT. In a NOTE before step 3.3.1 of OU-AA-200, and in the definition of RED SDR at step 5.3.20.d, the procedure OU-AA-200 clearly states, Planned entries into a RED condition are prohibited. This distractor is therefore incorrect in that, no matter how many compensatory actions are taken, planned entry into this condition is prohibited. Plausibility for this distractor is provided because the actions are correct for an unplanned entry into an ORANGE condition, with the added restrictions/concern over time in the RED SDR condition and the added restriction/concern over not allowing RED SDR entry if the plant is at mid-loop with fuel in the vessel.
B. INCORRECT. See analysis of A above.
C. INCORRECT. See analysis of A above.
D. CORRECT. No planned entry into a RED SDR condition is allowed by this procedure.
Supporting References
- 1. Dominion Administrative Procedure OU-AA-200, Shutdown Risk Management, rev. 0.
References Provided to Applicant None.
Answer: D
- 22. G2.2.7 1 Unit 1 is in Mode 3 and preparations are being made to conduct an infrequently performed test to verify the time dependence of reactor coolant flow following a loss of a reactor coolant pump.
Which one of the following states (1) whether this test is required to be designated as an Infrequently Conducted or Complex Evolution (ICCE) Category I or Category II in accordance with OP-AA-106, Infrequently Conducted or Complex Evolutions, and (2) an acceptable example of the individual who meets the general selection guidance for the seniority and qualifications of the Senior Operations Manager selected to provide oversight of the test in accordance with OP-AA-106.
A. (1) Category I (2) Control Room Supervisor with an active SRO license at Surry.
B. (1) Category I (2) Maintenance manager who formerly held an SRO license at Surry.
C. (1) Category II (2) Control Room Supervisor with an active SRO license at Surry.
D. (1) Category II (2) Maintenance manager who formerly held an SRO license at Surry.
K/A: G2.2.7 Knowledge of the process for conducting special or infrequent tests.
KA MATCH ANALYSIS:
The evolution in the stem comes directly from their procedure as an example of a Category I ICCE. The applicant must have knowledge that this test is a Category I and that a Manager or above with a current or past SRO must supervise the test.
SRO-ONLY ANALYSIS:
This task is purely an SRO function.
ANSWER CHOICE ANALYSIS:
Category I and Manager Level or above who formerly or currently held an SRO License is the correct answer. This information comes dierctly from AP-AA-106. The answer choices are plausible because it is credible that a current SRO license would be the more qualified supervisor. It is also plausible because an applicant must determine from memory whether it is a Cat I or II.
REFERENCES TO BE PROVIDED TO APPLICANT: None
REFERENCES:
- 1. OP-AA-106, Infrequently Conducted or Complex Evolutions, Rev. 4.
Answer: B
- 23. G2.3.14 1 Current conditions:
- Unit 1 is at full power.
Which one of the following completes the statement concerning:
- 1) the DOSE EQUIVALENT IODINE-131 limit for RCS activity in accordance with Technical Specification 3.1.D, Maximum Reactor Coolant Activity AND
- 2) the assumed release duration through the Main Steam Safety Valves and Atmospheric Relief Valves in accordance with Technical Specification bases?
The RCS activity must be limited to (1) DOSE EQUIVALENT IODINE-131. Primary water is assumed to enter the secondary system and be released for a period of (2) .
A. 1) Less than or equal to 0.1 µCi/cc.
- 2) 60 minutes.
B. 1) Less than or equal to 0.1 µCi/cc.
- 2) 30 minutes.
C. 1) Less than or equal to 1.0 µCi/cc
- 2) 60 minutes.
D. 1) Less than or equal to 1.0 µCi/cc
- 2) 30 minutes.
K/A Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.
(CFR: 41.12/43.4/45.11) (SRO - 3.8)
K/A Match Analysis The applicant must know the activity limits for normal full power operation the assumed duration for the release of radioactive/contaminated material to the public.
SRO Guidance Analysis The question requires the applicant to know the facility operating limitations in the Technical Specifications and their bases [(55.43 (b)(2)] and is beyond the expected knowledge requirements for an RO applicant.
Answer Choice Analysis A. INCORRECT. Plausible because this is the activity limit for the secondary coolant.
The second half of the response is actually twice the assumed release duration period of 30 minutes.
B. INCORRECT. Plausible because this is the activity limit for the secondary coolant.
The second half is the correct release duration period.
C. INCORRECT. Plausible because the first half of the response is the correct limit.
The second half of the response is actually twice the assumed release duration period of 30 minutes.
D. CORRECT. Per T.S 3.1.D.2 the limit is 1.0 µCi/cc and it based on a double-ended SGTR coincident with a loss of the main condenser. The release duration is assumed to last 30 minutes before the steam generator is manually isolated.
Supporting References
- 1. T.S. 3.1.D, Maximum Reactor Coolant Activity, pgs. 3.1-15a & 3.1-16
- 2. T.S. 3.1.H, Steam Generator Tube Integrity, (bases), pg.3.1-27
- 3. T.S. 3.6., Turbine Cycle, pgs. 3.6-4 & 3.6-5b
- 4. Modified question from Q# 98 on the Sequoyah 2009-302 [Retake Exam]. Modified
the question to ask for the RCS activity limit and the assumed release period that was the basis for the limit versus the regulatory section and SG tube leakage assumptions.
References Provided to Applicant None.
Answer: D
- 24. G2.4.29 1 Initial Conditions:
- Time = 0956. An automatic reactor trip and safety injection occurred on Unit 1, which had been operating at 100% power.
Current Conditions:
- Time = 1008. The Shift Manager (SM) announces entry into a Site Area Emergency (SAE).
Based on the current conditions, which one of the following is (1) the required elements of the Shift Managers announcement declaring the SAE, in accordance with EPIP-1.01, EMERGENCY MANAGER CONTROLLING PROCEDURE; and (2) if a General Emergency (GE) was declared, what are the time requirements associated with notifying outside agencies of the applicable Protective Action Recommendation (PAR), in accordance with EPIP-1.05, RESPONSE TO GENERAL EMERGENCY?
The Shift Manager is required to announce: that he or she has assumed the Station Emergency Manager (SEM) position, the emergency classification, the time of declaration, [AND] _________(1)____________ .
The initial notification of the applicable PAR _________(2)___________ .
A. (1) the EAL, and any pertinent upgrade criteria to a GE.
(2) is required to be made at the same time the initial notification of the GE is made.
B. (1) the EAL. Announcing upgrade criteria is not required by EPIP-1.01.
(2) is required to be made at the same time the initial notification of the GE is made.
C. (1) the EAL, and any pertinent upgrade criteria to a GE.
(2) is NOT required to be made at the same time the initial notification of the GE is made.
D. (1) the EAL. Announcing upgrade criteria is not required by EPIP-1.01.
(2) is NOT required to be made at the same time the initial notification of the GE is made.
K/A Generic: Emergency Procedures/Emergency Plan G2.4.29 Knowledge of the emergency plan.
(CFR 41.10 / 43.5 / 45.11) (SRO - 4.4)
K/A Match Analysis This question matches the K/A statement by requiring the SRO applicant to correctly determine the required elements of announcing entry into the emergency plan, and to correctly determine the required notifications associated with Protective Actions Recommendations (PAR).
SRO Guidance Analysis The question author intends to link this question to 10CFR55.43(b)(1) - conditions and limitations in the facility license. As a note, the SRO clarification guidance document does not include a flow chart for this topic. In this case, the author is linking the emergency plan as a part of the facility license. Only a SRO-qualified individual is qualified to assume the duties of the Site Emergency Manager (SEM); therefore, only a SRO is expected to know the portions of the E-plan that relate specifically to the SEM position. Therefore, the question is SRO-only.
Answer Choice Analysis A. INCORRECT. Part (1) of this distractor is incorrect; upgrade criteria is not required to be made as part of the announcement of the emergency. Announcing upgrade criteria is plausible because it is a common practice to alert the operations team to monitor key parameters that may trigger entry into a higher emergency classification; however, it is technically not required. EPIP-1.01 step 3 reads as follows:
____ 3 ANNOUNCE THE FOLLOWING DECLARATIONS:
- Station Emergency Manager position
- Emergency Classification
- Time Declared Part (2) of this distractor is correct. A NOTE before step 3 of EPIP-1.05 states: The initial notification of General Emergency and an applicable Protective Action
Recommendation (PAR) must be made to the State within 15 minutes following the declaration of the emergency. This NOTE is further reinforced by attachment 6 of EPIP-1.06, which states the following: The initial PAR must be included with the initial notification of a General Emergency, which must be made to the State within 15 minutes following declaration of the General Emergency.
B. CORRECT. See analysis of A above.
C. INCORRECT. Both part (1) and part (2) are incorrect; see analysis of A above. It is plausible to believe that the PAR notification may be separated from the initial notification of the GE, because the PAR has its own procedure, and it would be plausible, but incorrect, to believe that the SEM has extra time to determine the correct PAR, and could transmit the PAR separately from the GE notification.
D. INCORRECT. Part (1) is correct, part (2) is incorrect; see analysis of A and C above.
Supporting References
- 1. Surry Procedure EPIP-1.01, EMERGENCY MANAGER CONTROLLING PROCEDURE, rev. 52.
- 2. Surry Procedure EPIP-1.05, RESPONSE TO GENERAL EMERGENCY, rev. 21.
- 3. Surry Procedure EPIP-1.06, PROTECTIVE ACTION RECOMMENDATIONS, rev.
8.
References Provided to Applicant None.
Answer: B
- 25. G2.4.5 1 The following conditions exist:
- A manual Rx trip was initiated 10 minutes ago based on AP-16.00, Ecessive RCS Leakage, criteria
- Pressurizer level is off-scale low
- Pressurizer pressure is 1500 psig and decreasing
- All SG levels are 5% NR and slowly increasing
- All SG pressures are 1005 psig and stable
- All main steam line radiation monitors are reading .02 mr/hr
- MGPI Vent-Vent radiation monitor is reading 4.3 E6 cpm
- Containment pressure is 9.2 psia
- Containment sump level is 47%
- VSP-F-4, AUX Building Sump HI Level, is illuminated
- Safeguards Area Sump high level alarm is locked in Upon exiting E-0, which one of the following is the correct procedure transitions for the event in progress?
A. Go to E-1 (Loss of Reactor or Secondary Coolant), ECA-1.2 (LOCA Outside Containment), then ECA-1.1 (Loss of Emergency Coolant Recirculation)
B. Go to E-1 (Loss of Reactor or Secondary Coolant), then ECA-1.1 (Loss of Emergency Coolant Recirculation), the ECA-1.2 (LOCA Outside Containment)
C. Go to ECA-1.1 (Loss of Emergency Coolant Recirculation), then ECA-1.2 (LOCA Outside Containment)
D. Go to ECA-1.2 (LOCA Outside Containment), then ECA-1.1 (Loss of Emergency Coolant Recirculation)
K/A Knowledge of the organization of the operating procedures network for normal, abnormal, and emergency evolutions.
K/A Match Analysis The question requires knowledge of how the different subsets of emergency procedures are used in conjunction with each other to combat an event.
SRO Guidance Analysis Knowledge required of 10 CFR 55.43 is required due to procedure selection. The procedure selection knowledge required by this question is beyond the entry conditions of major EOPs and FRPs; thereby testing the procedure selection at the SRO level.
Answer Choice Analysis A. INCORRECT. see D below. Plausible because E-1 combats LOCAs and the other two procedures are part of the correct answer.
B. INCORRECT. see D below. Plausible because E-1 combats LOCAs and the other two procedures are part of the correct answer.
C. INCORRECT. see D below. The procedures are part of the correct answer, but in this choice they are in the wrong sequence.
D. CORRECT. With a LOCA outside containment that cannot be isolated, the correct transitions following E-0 are as stated in this choice. See supporting copy of E-0 which shows the correct flow path through ECA-1.2 to ECA-1.1.
Supporting References AP-16.00, Ecessive RCS Leakage, Rev 17.
1-E-0, Reactor Trip or Safety Injection, Rev. 63.
1-E-1, Loss of Reactor or Secondary Coolant, Rev 34.
1-ECA-1.2, LOCA Outside Containment, Rev. 7.
References Provided to Applicant None.
Answer: D