ML23346A099
| ML23346A099 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 11/16/2023 |
| From: | -Neednewvalue Wolf Creek |
| To: | Kelly Clayton NRC Region 4 |
| References | |
| Download: ML23346A099 (1) | |
Text
DATE: November 16, 2023 TO: Kelly Clayton (USNRC)
FROM: Andrew Servaes (Evergy - Wolf Creek)
SUBJECT:
Wolf Creek Generating Station Written Exam Results and Post Exam Items for 2023 Initial License Examination.
Dear Mr. Clayton,
Enclosed are the Written Examination Results and Post Exam Analysis and all Post Exam items required by NUREG 1021, Paragraph ES 4.4.B.7. Please ensure that all materials associated with this examination are withheld from public disclosure for a two-year period ending in November 2025. Of note, this was the first examination which incorporated Tier 4 Fundamental questions. Specific analysis for these questions is as follows [90.9% average (60/66)]:
- Q70 - Reactor Theory - Neutron Life Cycle - 100% (11/11)
- Q71 - Reactor Theory - Reactivity Coefficients - 90.9% (10/11)
- Q72 - Reactor Theory - Control Rods - 100% (11/11)
- Q73 - Thermodynamics - Steam - 72.7% (8/11)
- Q74 - Thermodynamics - Throttling and the Throttling Process - 100% (11/11)
- Q75 - Thermodynamics - Core Thermal Power - 81.8% (9/11)
Included in this post exam package are:
- 1. The graded written examination results.
- a. (11) Original ungraded Scantron forms.
- b. Blank Written Exam Cover Sheets (5) Form 4.3-3, Reactor Operator and (6) Form 4.3-4, Senior Reactor Operator.
- a. As-given Exams and Original Answer Key.
- b. Original Answer Key Scantron Template.
- 3. (3) Question clarifications asked during written exam administration on 11/1/2023 as documented on station form AIF 30B-015-14 for Q16, Q20, and Q90.
- 4. All examination administration and post examination review comments including the results of written exam performance analysis and recommended substantive changes.
- a. Written Exam Question Performance, report generated from VISION program.
- b. Post Exam Analysis for High Miss and flawed question documented on station form NISP TR-01 FORM 9. Details of (5) flawed questions documented below. (2) questions were missed by more than 50% of the applicants, but were determined to be acceptable, and the knowledge gap was closed by post-exam debrief.
Q29 - (27.3% - 3/11) - Rod Control System (001/A2.14)
Q80 - (33.3% - 2/6) - Service Water System (076/A2.04)
- c. (5) Question challenges documented on station form AIF 30B-015-13.
- d. Regrade Requests for the following (5) Questions after considering question challenges and additional information. The issues associated with these questions were not recognized during exam development or validation.
Q33 - Change correct answer to A (from C).
Q41 - Remove flawed question from exam.
Q49 - Accept D as second correct answer (with A).
Q57 - Remove flawed question from exam.
Q63 - Accept B as second correct answer (with C).
- e. Additional References supplied for regrade request consideration:
Q33 - Westinghouse Proprietary Class 2 - TC/CCM Operations and Maintenance Manual, Wolf Creek Drawing M-764A-00045.
Q49 - IRIS Report 252550, Wolf Creek Operating Experience Report from 1/13/12.
Q57 - Original Bank Question (LO19739).
Q57 - SBT Package and Simulator Data to support removal from exam.
Q63 - Bank Question (LO116353) version from last use in 2015.
- f. (3) Post Examination Condition Reports.
CR10028338, General Weakness (Action Verbs) for Simulator JPM failures.
CR10028341, General Weakness (MFP SLIM Controller) for Simulator JPM failures.
CR10028344, ILO Written Examination Preliminary Results and flawed questions.
- g. Summary of Post Examination Lesson Plan (6) and Procedure (2) Enhancement suggestions.
- 5. Seating Chart for the Written Examination.
- 6. Form 1.3-1, Examination Security Agreement.
Sincerely, Andrew J. Servaes Exam Author James P. Knapp Facility Representative
Recommended Changes to the 2023 Wolf Creek NRC Initial Written Exam.
Original Question:
33 ID: 161863 Points: 1.00 With the unit operating at 100% power, one of the In-Core Thermocouples failed due to an open circuit.
Based on these conditions, the NPIS computer display for the failed thermocouple will...
A.
indicate 0°F.
B.
indicate 2300°F.
C.
be blank with flashing magenta field.
D.
indicate the calibrated reference junction temperature.
Answer:
D Answer Explanation (Recommend changing the Correct Answer to A)
Answer choices sorted. Correct answer at D.
A. Distractor 1 (indicate 0°F.) is INCORRECT, but plausible. Even though an open circuit will cause a thermocouple to fail in the low direction, the indication will match the calibrated reference junction temperature. Per WCRE-01, WCNOC TOTAL PLANT SETPOINT DOCUMENT, the T/C Junction Box Cold Junction Temperature Setpoint range is 32-420°F.
Incore Thermocouples setpoint range is 0-2300°F. This choice is plausible since an RTD will read zero on a short circuit. This answer choice was selected as the correct answer by a Licensed Operator during validations.
B. Distractor 2 (indicate 2300°F.) is INCORRECT, but plausible. An RTD will read high scale on an open circuit, but thermocouples fail to the calibrated reference junction temperature, on either a short or open circuit failure. This is one of the differences between a thermocouple and an RTD. This answer choice was selected as the correct answer by a Licensed Operator during validations.
C. Distractor 3 (be blank with flashing magenta field.) is INCORRECT, but plausible. Flashing magenta is an indication of a Loss of NPIS Computer, but this choice is wrong for a failure of an input to NPIS. This answer choice was selected as the correct answer by a Licensed Operator during validations.
D. CORRECT (indicate the calibrated reference junction temperature.) Two dissimilar metals, when heated, will develop a voltage when one junction is heated relative to the cold junction. If the junction between the dissimilar metals is interrupted by an open circuit, no path for current flow exists, and thus the temperature indication will fail low. Keep in mind that a thermocouple also employs a reference junction. When this junction is calibrated to some temperature above 0°F, depending on where the open exists, the thermocouple would fail to the reference junction calibrated temperature (which is still in the low direction).
Recommended Changes to the 2023 Wolf Creek NRC Initial Written Exam.
Question 33 Info Time to Complete:
2 Difficulty:
2.00 System ID:
161863 User-Defined ID:
(23) LO161862
Reference:
ITMS Topic:
- 33 - 017 / K3.02 (New) - N/F/2/D RO Importance Rating:
3.2 SRO Importance Rating:
CFR: 41.02 K/A Number:
017 K3.02 Comments:
2023 Wolf Creek NRC Exam, K/A Statement, In-Core Temperature Monitor System (ITMS), Knowledge of the effect that a loss or malfunction of the ITMS will have on the following systems or system parameters: Plant Computer.
Tier 2 Group 2 Fundamental Technical
References:
LO1301700, Incore Nuclear Instrumentation System, Objective 4, Explain the operation of the Incore Thermocouple System major components (Page 27).
Sensors and Detectors, Part 1 Student Guide. ELO 1.2 Environmental Effects/ Detector Failures (Page 17).
WCRE-01 WCNOC Total Plant Setpoint Document (Page 280).
OFN RJ-023, NPIS MALFUNCTIONS (Page 2).
Meets Tier 2 Criteria as question is focused on knowledge of plant systems, components, and/or their interrelations.
Meets the K/A since the question presents an In-Core Temperature Monitor malfunction (failed in-core thermocouple) and asks for the response as indicated on the plant computer.
Question History:
Rev 0 - Submitted for 2023 Wolf Creek NRC Exam.
Recommended Changes to the 2023 Wolf Creek NRC Initial Written Exam.
New Information that supports question regrade request:
Applicants were briefed per NUREG 1021, ES1.2, paragraph 8 Finally, answer all questions based on actual plant operation, procedures, and references. If you believe that the answer would be different based on simulator operation or training references, you should answer the question based on the actual plant.
In the actual plant, we have four failed Core Exit Thermocouples (CETCs) - see grid locations J-10, G-06, G-08, and N-02. This display is provided to show CETC location and not what would be observed on the plant computer for a spuriously failed Open condition.
From the plant computer, actual data for these failed CETCs is as follows:
See Note 1 on display above, CETC J-10 is physically removed, while G-06 and G-08 are Abandoned in place. N-02 was removed from service during startup from RF25 on 11/13/2022 and the following Equipment out of service log entry was made.
Recommended Changes to the 2023 Wolf Creek NRC Initial Written Exam.
A summary of the associated work order for CETC N-02 (WR 22-144211) is as follows:
The cause of the failure for CETC (N-02) was not an open, and the indication did not fail to 0.0°F, nor the calibrated Reference Junction Temperature value. The shown 257.297°F value in Magenta color represents the last known value when the individual CETC input to the Core Monitoring System was removed from scan. This data point discussion for CETC (N-02) is provided since we initaly considered that value to possibly be the Calibrated Reference Junction Temperature.
A search was performed for previous examples of failed open thermocouples and one (Location E-06) was found during Fuel Cycle 13 in 2002.
Recommended Changes to the 2023 Wolf Creek NRC Initial Written Exam.
The associated value for that CETC (E-06) during the time frame where the documented Open existed was 0.0°F as shown on this graph below for the time period between 4/28/2002 and January 5, 2003.
This is the most complelling evidence that supports the challenge that 0.00°F is the right answer and only right answer.
As part of the refueling process, each CETC is disconnected to support head removal. In each case, this break in connectivity represents an Open in the circuit and the expected Plant Computer indication for this Open Conditions is 0.0°F as demonstrated by this trend/graph from last outage (RF25) in 2022.
Recommended Changes to the 2023 Wolf Creek NRC Initial Written Exam.
Finally, drawing M-764A-00045 (Westinghouse Proprietary, Class 2) Figure 2.1-2 (on page 20), shows the inputs to the TC/CCM System in which CETC is a separate input (in parallel, not series) with the Reference Junction Temperature RTD input. In all observed cases of Thermocouple failure due to an Open or disconnected condition, the expected plant computer indication is 0.0°F.
The drawing listed above has been removed from this documentation because of its proprietary classification.
Exam Author Perspective:
When I wrote the question, I believed D was the only correct answer based on fundamental (GFES) knowledge of how a Thermocouple circuit fails upstream of the Calibrated Reference Junction in a simplified theoretical detector. I wrote the question based on NRC Bank GFES question P213, but I removed the word detector and changed the four answer choices make the question operationally valid without recognizing or applying actual system or plant design.
This exam construction error is documented on CR10028344 which will be used to establish corrective actions that will prevent recurrence on future exams. Along with #57, I made errors while trying to test a fundamental concept in an Operationally valid manner. Bank question will be updated.
Recommendation:
Based on actual plant data and system design, A is the only correct answer as an Open or disconnected CETC will result in an indication of 0.000 DEG F on the plant computer.
Original Question:
Recommended Changes to the 2023 Wolf Creek NRC Initial Written Exam.
41 ID: 161954 Points: 1.00 Given:
An earthquake caused a loss of off-site power and a Large Break LOCA in containment.
In accordance with EMG FR-C2, RESPONSE TO DEGRADED CORE COOLING, the crew is depressurizing S/Gs to 250 psig.
Based on these conditions, and per EMG FR-C2,
- 1) How will the crew perform the S/G depressurization?
- 2) At what rate will the cooldown be performed?
A.
- 1) Using ARVs
- 2) at Maximum Rate B.
- 1) Using ARVs
- 2) at less than 100°F/hr C.
- 1) Using Steam Dumps
- 2) at Maximum Rate D.
- 1) Using Steam Dumps
- 2) at less than 100°F/hr Answer:
B Answer Explanation (Recommend removal from exam)
Answer choices sorted. Correct answer at B.
A. Distractor 1 (ARV, Max Rate) is INCORRECT, but plausible. This choice would be right if the crew were responding to INADEQUATE core cooling per EMG FR-C1. Wrong since EMG FR-C2, step 15 directs maintaining a controlled cooldown at a rate <100F/hr. This answer choice was selected as the correct answer by a Licensed Operator during validations.
B. CORRECT (ARV, 100F/hr). Per EMG FR-C2 and for the given conditions (LOOP), Steam dumps are NOT available, so Step 15 directs manually dumping steam from S/Gs using S/G ARV awhile maintaining cooldown rate in RCS cold legs <100F/hr.
C. Distractor 2 (Steam Dumps, Max Rate) is INCORRECT, but plausible. Both answer choices are wrong; the opposite of the correct answer.
D. Distractor 3 (Steam Dumps, 100F/hr) is INCORRECT, but plausible. This choice would be right if Steam Dumps were available, but is wrong since without off-site power, Steam dumps are NOT available without any available circulating water pumps (C-9). The cooldown rate is right.
Recommended Changes to the 2023 Wolf Creek NRC Initial Written Exam.
Question 41 Info Time to Complete:
3 Difficulty:
3.00 System ID:
161954 User-Defined ID:
(23) LO161954
Reference:
- 41 - EPE 011 / EA2.10 (100%) - B/CA/3/B RO Importance Rating:
4.6 SRO Importance Rating:
CFR: 41.10 K/A Number:
EPE 011 EA2.10 Comments:
2023 Wolf Creek NRC, K/A Statement, Large Break LOCA, Ability to determine and/or interpret the following as they apply to a Large-Break LOCA: Adequate core cooling.
Tier 1 / Group 1 / Comprehension / Analysis Technical
References:
LO1732341, EMG FR-C1/C2/C3 Inadequate/Degraded/Saturated Core Conditions, Explain the basis and any knowledge requirements for procedure steps of EMG FR-C2 (Page 22).
EMG FR-C2, RESPONSE TO DEGRADED CORE COOLING (Step 15, Page 24).
Question meets Tier 1 requirements since it tests an applicants knowledge of how to safely operate the plant during emergency and abnormal conditions. This objective is met using:
(1) information contained in the sites procedures, including emergency operating procedures (EOPs), and their associated bases documents, (2) diagnosis that leads to selection of the procedures that should be used to respond to the evolution, (3) the progression of an event; and (4) the assessment of the integrated plant response to emergency or abnormal situations crossing several plant systems or safety functions, or both.
Meets the K/A at the RO Level since the stem presents a Large Break LOCA with less than adequate core cooling conditions and asks how the crew will depressurize the Intact S/Gs and at what rate to address this lack of core cooling. One of the major action categories of procedure EMG FR-C2 is to initiate a controlled S/G depressurization to cool down and depressurize the RCS, which will enable the SI Accumulators to inject and cover the core. Overlap with #7 considered to focus the question on secondary side response and not focus on ECCS.
Overlap with #62 also considered, as that Inadequate Core Cooling Topic tests RCP operation per EMG FR-C1, based on CETC temperatures.
Question History:
Bank Question (Copy of LR153245)
Rev 0 - Submitted for 2023 Wolf Creek NRC Exam
Recommended Changes to the 2023 Wolf Creek NRC Initial Written Exam.
New Information that Supports removal of question from the exam:
The applicants were briefed per NUREG 1021, ES1.2, paragraph 8 When answering a question, do not make assumptions about conditions that are not specified in the question unless they occur as a consequence of other conditions that are stated in the question For the given stem conditions - Large Break LOCA with no equipment out of service, RCS Pressure is already below SI Accumulator Injection pressure and full ECCS flow from two safety trains will be providing more than adequate cooling (in excess of 100F//hr) and cooldown rate will not be under Operator control. For these given conditions, the crew will not enter EMG FR-C2, RESPONSE TO DEGRADED CORE COOLING, which makes this question invalid.
Entry conditions for EMG FR-C2 might be possible with a series of equipment failures, but during a Large Break LOCA, the mitigation strategy provided by Step 15 to bleed steam to reduce RCS Pressure in order for the SI accumulator and RHR pumps to inject is already accomplished by the size of the RCS break.
EMG FR-C2 step 15 Basis -
BASIS:
The controlled secondary depressurization, similar to the one in EMG ES-11, POST LOCA COOLDOWN AND DEPRESSURIZATION, has been shown to be an effective way to reduce RCS pressure. RCS pressure must be reduced in order for the SI accumulator and RHR pumps to inject. To prevent accumulator nitrogen injection, the operator should stop the secondary depressurization when the S/G pressure reaches 250 psig.
As a result of the given Large Break LOCA, RCS pressure will be below Steam Generator pressure which de-couples the RCS from the secondary. EMG FR-H1, RESPONSE TO LOSS OF SECDARY HEAT SINK, Step 1 and the basis for this step provide a good description of why entry to EMG FR-H1 is not required during a Large Break LOCA - the S/Gs no longer function as a heat sink and heat removal will be accomplished by RCS break flow.
EMG FR-H1 Step 1 Basis -
BASIS:
Before implementing actions to restore flow to the S/Gs, the operator should check if secondary heat sink is required. For larger break LOCAs, the RCS will depressurize below the intact S/G pressures. The S/Gs no longer function as a heat sink and the core decay heat is removed by the RCS break flow. For this range of LOCA break sizes, the secondary heat sink is not required and actions to restore secondary heat sink are not necessary.
Exam Author Perspective: I selected a degraded core cooling question from the bank. That bank question didnt specify a LOCA size, so I added Large Break to strengthen the tie to the given K/A, which inadvertently invalidated the question. This exam construction error is documented on CR10028344 which will be used to establish corrective actions that will prevent recurrence on future exams. The bank question will be edited to specify Small Break and a different K/A (EPE 009) will be assigned.
Recommendation: Remove this invalid question from the exam.
Recommended Changes to the 2023 Wolf Creek NRC Initial Written Exam.
Original Question:
49 ID: 161963 Points: 1.00 Given:
A Loss of Off-Site Power occurred.
MCB Annunciator 014A, S/U XFMR LOCKOUT actuated when the unit tripped.
The crew was responding per OFN NB-030, LOSS OF AC EMERGENCY BUS NB01 (NB02) when Off-Site power was restored to the switchyard.
Based on these conditions, which of the following actions should the crew take as directed by OFN NB-030?
A.
Transfer NB01 power supply from DG NE01 to Normal FDR BKR.
B.
Transfer NB01 power supply from DG NE01 to Alternate FDR BKR.
C.
Transfer NB02 power supply from DG NE02 to Normal FDR BKR.
D.
Transfer NB02 power supply from DG NE02 to Alternate FDR BKR.
Answer:
A Answer Explanation (Recommend accepting both A and D as Correct Answers)
Answer choices sorted. Correct Answer at A.
A. CORRECT (Transfer NB01 power supply from DG NE01 to Normal FDR BKR. OFN NB-030, Step A42b RNO directs the crew to try to restore offsite power to NB01, using SYS NB-201, TRANSFERRING NB01 POWER SOURCES. SYS NB-201 provides direction to transfer NB01 power from A EDG to either Normal or Alternate feeder breakers. AC Distribution System knowledge is required to determine power from alternate feeder breaker is unavailable given Startup Transformer lockout.
B. Distractor 1 (Transfer NB01 power supply from DG NE01 to Alternate FDR BKR.) is INCORRECT, but plausible. OFN NB-030, Step A42b RNO directs the crew to try to restore offsite power to NB01, using SYS NB-201, TRANSFERRING NB01 POWER SOURCES. SYS NB-201 provides direction to transfer NB01 power from A EDG to either Normal or Alternate feeder breakers. AC Distribution System knowledge is required to determine power from alternate feeder breaker is unavailable given Startup Transformer lockout.
C. Distractor 2 (Transfer NB02 power supply from DG NE02 to Normal FDR BKR.) is INCORRECT, but plausible. OFN NB-030, Step B42b RNO directs the crew to try to restore offsite power to NB02, using SYS NB-202, TRANSFERRING NB02 POWER SOURCES. SYS NB-202 provides direction to transfer NB02 power from B EDG to either Normal or Alternate feeder breakers. AC Distribution System knowledge is required to determine power from normal feeder breaker is unavailable given Startup Transformer lockout. This answer choice was selected as the correct answer by a Licensed Operator during validations.
D. Distractor 3 (Transfer NB02 power supply from DG NE02 to Alternate FDR BKR.) is INCORRECT, but plausible. OFN NB-030, Step B42b RNO directs the crew to try to restore offsite power to NB02, using SYS NB-202, TRANSFERRING NB02 POWER SOURCES. SYS NB-202 provides direction to transfer NB02 power from B EDG to either Normal or Alternate feeder breakers. NB02 could physically be energized from the Alternate Feeder breaker and this lineup might be considered if a NB01 bus lockout also existed coincident with the loss of Startup Transformer. NB01 and NB02 may not be cross connected from the same power source. This answer choice was selected as the correct answer by a Licensed Operator during validations.
Recommended Changes to the 2023 Wolf Creek NRC Initial Written Exam.
Question 49 Info Time to Complete:
3 Difficulty:
3.00 System ID:
161963 User-Defined ID:
(23) LO161963
Reference:
LOOP Topic:
- 49 - APE 056 / AK2.10 (New) - N/CA/3/A RO Importance Rating:
4.3 SRO Importance Rating: CFR: 41.10 K/A Number:
APE 056 AK2.10 Comments:
2023 Wolf Creek NRC Exam, K/A Statement, Loss of Off-Site Power (LOOP), Knowledge of the relationship between the LOOP and the following systems or components: AC Distribution system.
Tier 1 / Group 1 / Comprehension / Analysis Technical
References:
LO1732444, OFN NB-030, Loss of AC Emergency Bus NB01 (NB02) and OFN N-042, LOOP with EDG Paralleled (Page 19).
LO1506205, Power Block AC Electrical Distribution, Objective 7, Describe the system and plant response to Power Block AC electrical distribution system failures (Pages 47 and 48).
OFN NB-030, LOSS OF AC EMERGENCY BUS NB01 (NB02)
(Step A42, Page 64).
SYS NB-201, TRANSFERRING NB01 POWER SOURCES (Section 6.4, Page 39).
KD-7496, ONE LINE DIAGRAM (E-8).
Question meets Tier 1 requirements since it tests an applicants knowledge of how to safely operate the plant during emergency and abnormal conditions. This objective is met using:(1) information contained in the sites procedures (SYS NB-201), including alarm response procedures, abnormal operating procedures (OFN NB-030),
and their associated bases documents (3) the progression of an even (restoration of off-site power)t; and (4) the assessment of the integrated plant response to emergency or abnormal situations crossing several plant systems or safety functions, or both (AC Distribution System, EDGS).
Meets the K/A since the stem presents a loss of off-site power scenario with a loss of startup transformer and the question tests restoration of the AC Distribution System lineup as directed by OFN NB-030.
Question History:
Rev 0 - Submitted for 2023 Wolf Creek NRC Exam.
Rev 1 - Updated cognitive level based on NRC Feedback.
Recommended Changes to the 2023 Wolf Creek NRC Initial Written Exam.
New Information that Supports regrade request:
The applicants were briefed per NUREG 1021, ES1.2, paragraph 8 When answering a question, do not make assumptions about conditions that are not specified in the question unless they occur as a consequence of other conditions that are stated in the question Finally, answer all questions based on actual plant operation, procedures, and references. If you believe that the answer would be different based on simulator operation or training references, you should answer the question based on the actual plant.
The given conditions do not specify what MODE the unit is in or how much time elapsed between the loss of off-site power and the restoration of off-site power. With the given Start-Up Transformer lockout, the crew has no timely option for restoration of off-site power to the non-safety related buses. Therefore, it is reasonable to assume the crew would quickly move to initiate a natural circulation cooldown to enter MODE 5 for the given condition. For the given Start-Up Transformer Lockout, Answer choices B and C are clearly wrong since power is unavailable. OFN NB-030 provides guidance and flexibility to allow the crew (SRO discretion) to determine preferred restoration lineup and the CRS would likely choose to energize BOTH NB01 and NB02 from the available safety related transformer once the unit reached MODE 5. For the given stem conditions, answer choice D is a second correct answer if the student applies the expected ES 1.2, paragraph 8 standard to determine the crew took action to place the unit in MODE 5 and energized both NB01 and NB02 busses from XNB01 transformer.
Wolf Creek Operating Experience (IRIS #252550), Loss Of Off-Site Electrical Power (LOOP)
Following Unit Trip is applicable. On January 13, 2012, the station experienced a catastrophic failure of the startup transformer following an automatic reactor trip due to switchyard breaker failures. Both NB01 and NB02 were powered from associated safety-related Emergency Diesel Generators (EDG) until the switchyard east bus was re-energized and the crew transferred safety related bus NB01 power from A EDG to XNB01 Normal source. The crew performed a natural circulation cooldown to MODE 5 and eventually transferred safety related bus NB02 power from B EDG to XNB01 Alternate source since the Startup Transformer was out of service until February 13, 2012. This actual plant lineup and Operating Experience supports D as a second Correct Answer since MODE and timeframe to restore power to the switchyard were not specified in the stem.
OFN NB-030, LOSS OF AC EMERGENCY BUS NB01 (NB02), Step B42:
SYS NB-202, TRANSFERRING NB02 POWER SOURCES, Precaution and Limitation 5.3.
Busses SL3 and SL4 are NOT cross-tied and are both de-energized for the given loss of off-site power.
SYS NB-202, Section 6.5 Transfer from DG NE02 to Alternate FDR BKR would be used by the crew to accomplish this electrical line-up.
Recommended Changes to the 2023 Wolf Creek NRC Initial Written Exam.
Exam Author Perspective:
When I wrote this question, I believed A to be the only correct answer as I considered the unit had been operating at 100% power prior to the loss of Offsite power and restoration of the off-site power occurred while the unit remained in MODE 3. Even though I had a Licensed Operator select D during the validation process, we discussed and determined the crew would not be procedurally allowed to establish an alternate lineup unless in MODE 5. I failed to update the answer explanation definitively to state why D was wrong in MODE 3, nor specify in the stem the timeframe in which off-site power was restored, to make it clear the unit remained in MODE 3. This exam construction error is documented on CR10028344 which will be used to establish corrective actions that will prevent recurrence on future exams. The bank question will be updated to specify the unit is in MODE 3 or power was restored one hour later, and the answer explanation will be updated to specify why D is wrong.
Recommendation:
Since the question stem failed to specify the unit MODE and the time frame in which off-site power was restored, accept both A and D as correct answers.
Recommended Changes to the 2023 Wolf Creek NRC Initial Written Exam.
Original Question:
57 ID: 162041 Points: 1.00 Given:
Reactor power is steady at 70% for FLEX operations.
Rod Control is in MANUAL.
Ovation is in First Stage Pressure Mode.
All other control systems are in automatic.
In response to equipment malfunction, the crew performed an Emergency Boration per OFN BG-009, EMERGENCY BORATION for two minutes.
Assuming NO other Operator actions, which parameter will return closest to its pre-boration value after steady state conditions are attained?
A.
PZR Level B.
RCS Tavg C.
S/G Pressure D.
Reactor Power Answer:
D Answer Explanation (Recommend removal from exam)
Answer choices sorted. Correct answer at D.
A. Distractor 1 (PZR Level) is INCORRECT, but plausible. PZR level is based on Tavg. A lower Tavg will lower the setpoint and control level at a lower value. This answer choice was selected as the correct answer by a Licensed Operator during validations.
B. Distractor 2 (RCS Tavg) is INCORRECT, but plausible. RCS Tavg lowers due to more poison and its change in thermal neutron utilization (p). With a -MTC this causes temperature to lower, raising moderator density to raise resonance escape probability (f) to offset the higher capture of thermal neutrons.
C. Distractor 3 (S/G Pressure) is INCORRECT, but plausible. S/G Pressure is based on saturation conditions in the S/G and loop hot leg / cold leg values. Hot leg temperatures will lower causing S/G pressure to lower with it as well. This answer choice was selected as the correct answer by a Licensed Operator during validations.
D. CORRECT (Reactor Power) Reactor power is a function of steam load and since the turbine admission valve did not change, reactor power returns to ~70% at a lower MWE output. OFN BG-009, Note prior to step 1 specifies Prolonged Emergency Boration with the plant at power may require the plant to be tripped due to rapid RCS temperature, pressure and PZR Level drops.
Recommended Changes to the 2023 Wolf Creek NRC Initial Written Exam.
Question 57 Info Time to Complete:
3 Difficulty:
3.00 System ID:
162041 User-Defined ID:
(23) LO162041
Reference:
EMERGENCY BORATION Topic:
- 57 - APE 024 / AK1.02 (61%) - B/CA/3/D RO Importance Rating:
4.1 SRO Importance Rating:
CFR: 41.06 K/A Number:
APE 024 AK1.02 Comments:
2023 Wolf Creek NRC Exam, K/A Statement, Emergency Boration, Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to Emergency Boration: Relationship between boron addition and reactor power.
Tier 1 Group 2 Comprehension / Analysis Technical
References:
LO1732419, OFN BG-009, EMERGENCY BORATION, Objective 4: Explain the basis and knowledge requirement for selected procedure steps (Page 8).
OFN BG-009, EMERGENCY BORATION (Page 3).
Question meets Tier 1 requirements since it tests an applicants knowledge of how to safely operate the plant during emergency and abnormal conditions. This objective is met using: (1) information contained in the sites procedures, including alarm response procedures, abnormal operating procedures (OFN BG-009), and their associated bases documents, (3) the progression of an event; and (4) the assessment of the integrated plant response to emergency or abnormal situations crossing several plant systems or safety functions, or both.
Meets the K/A since the question presents a scenario where the crew performed an Emergency Boration at power and tests the relationship between that boration addition and reactor power.
Question History:
Bank Question (Copy of LO19739, with previous use on 2002 Callaway NRC Exam).
Rev 0 - Submitted for 2023 Wolf Creek NRC Exam
Recommended Changes to the 2023 Wolf Creek NRC Initial Written Exam.
New Information that Supports removal of question from the exam:
The applicants were briefed per NUREG 1021, ES1.2, paragraph 8 When answering a question, do not make assumptions about conditions that are not specified in the question unless they occur as a consequence of other conditions that are stated in the question Finally, answer all questions based on actual plant operation, procedures, and references. If you believe that the answer would be different based on simulator operation or training references, you should answer the question based on the actual plant.
We ran the given scenario on the full scope simulator in both OPEN LOOP (without Main Turbine Control feedback mechanism) and in FSP LOOP (First Stage Pressure Loop) to determined expected response for the four given generic parameters in each method of selected Main Turbine Ovation Control.
With Main Turbine Controls in OPEN LOOP, all four parameters lowered to a new equilibrium value. The boration lowered Tavg, which also lowered program PZR Level, and resulted in lowering S/G Pressure.
As a result of fixed Main Control Valve position and lower S/G pressure, steam flow lowered to the Main turbine resulting in a new lower equilibrium Reactor Power condition, both by Nuclear Instrument and Calorimetric - Core Thermal Power indications. Actual PZR level cycled due to system response adjusting charging flow to restore level to new lower program level.
2400 2420 2440 2460 2480 2500 2520 2540 2560 0
01:21.0 02:43.0 04:05.0 05:27.0 06:49.0 08:11.0 09:33.0 10:55.0 12:17.0 13:39.0 15:01.0 16:23.0 17:45.0 19:07.0 20:29.0 21:51.0 23:13.0 24:35.0 25:57.0 27:19.0 28:41.0 30:03.0 31:25.0 32:47.0 Core Thermal Power (MWt)
Recommended Changes to the 2023 Wolf Creek NRC Initial Written Exam.
With Main Turbine Controls in FSP LOOP, the Main Control Valves opened slightly to maintain a constant First Stage Pressure which resulted in similar response for Tavg, S/G Pressure, and PZR Level, but the Reactor Power Indication by Nuclear Instruments did not lower as much as when OPEN LOOP was selected. The Reactor Power indication by Nuclear Instruments also did not Return from new lower equilibrium value, so selecting which parameter remained closest to the pre-boration value becomes a subjective exercise. The Calorimetric - Core Thermal Power (MWt) value returned to approximately the same pre-boration value, so this demonstrates the simulator modeled fundamental theory. Reactor Power could have been defended as if the answer choice had specified Reactor Power as indicated by Calorimetric - Core Thermal Power (MWt). However, the indicated Reactor Power on the Nuclear Instruments remained at a lower value as that indication was affected by the additional boron and effects of lower temperature. It is for this reason, the crew will perform STS SE-001, POWER RANGE ADJUSTMENT TO CALORIMETRIC to calibrate the less accurate Reactor Power Indication to match the calculated thermal power after plant conditions are stable.
2505 2510 2515 2520 2525 2530 2535 2540 0
01:14.0 02:29.0 03:44.0 04:59.0 06:14.0 07:29.0 08:44.0 09:59.0 11:14.0 12:29.0 13:44.0 14:59.0 16:14.0 17:29.0 18:44.0 19:59.0 21:14.0 22:29.0 23:44.0 24:59.0 26:14.0 27:29.0 28:44.0 29:59.0 31:14.0 32:29.0 Core Thermal Power (MWt)
Recommended Changes to the 2023 Wolf Creek NRC Initial Written Exam.
Comparison of Key Parameters in each data run:
OPEN LOOP Answer Choice Parameter Initial Final A. PZR Level 47.5%
43.2%
4.3%
9.1%
B. Auct Hi Tavg 577.2°F 572.0°F 5.2°F 0.9%
C. S/G Pressure 984.0 psig 948.5 psig 35.5 psig 3.6%
D. Reactor Power (by NIs) 69.4%
64.7%
4.7%
6.7%
Control Valve Position 30.3%
30.3%
Core Thermal Power 2537.67MWt 2455.94 MWt 81.9 MWt 3.2%
FSP LOOP Answer Choice Parameter Initial Final A. PZR Level 47.5%
42.3%
5.2%
10.9%
B. Auct Hi Tavg 577.2°F 570.8°F 6.4°F 1.1%
C. S/G Pressure 984.0 psig 930.9 psig 53.1 psig 5.4%
D. Reactor Power (By NIs) 69.4%
66.5%
2.9%
4.2%
Control Valve Position 30.3%
31.4%
Core Thermal Power 2537.67 MWt 2535.69 MWt 1.98 MWt 0.1%
Applicants were also confused by the compelling need to Emergency Borate at power, and by what Equipment Malfunction would result in the crews need to perform that Emergency Boration. The ambiguity added by that stem statement made each applicant speculate the nature of the equipment malfunction and whether or not the results might be impacted by that arbitrary equipment malfunction statement. OFN BG-009 lists symptoms/entry conditions that might be required at power include unexplained or uncontrolled reactivity rise as indicated by abnormal control bank insertion (not given -
rods in manual), rising temperature or nuclear power (not given - Steady at 70%). There was no operational need to perform an emergency boration for the given conditions, so the question should have tested what would happen if instead of trying to focus on making the question operationally valid.
The question attempted to test fundamental knowledge in an operationally valid way, but the answer choices were not specific and since each parameter changed, the correct answer is based on subjectivity when comparing initial to final values with different units. How does 2.9% change by Nuclear Instrument indication compare in magnitude to a change of 53.1 psig, 6.4°F, or 5.2% level? Using a percentage of change from initial value, a case could be made that Auctioneered Hi Tavg is the closest to pre-boration value at 1.1% lower, compared to 4.2% lower NI reading, 5.4% lower S/G Pressure, and ~10.9% lower PZR level.
Exam Author Perspective:
I selected a bank question that that matched the given K/A and tested a fundamental concept in generic terms. In my attempt to make the question operationally valid, I inadvertently changed the complexity of the question by inserting an arbitrary reason for why the crew might be performing OFN BG-009 at power and I also failed to change the generic parameters in the answer choices. Reactor Power indications are diverse (Nuclear Instrument, Loop T, and Calorimetric - Core Thermal Power) and these indications have differences in accuracy and response as highlighted above. This exam construction error is documented on CR10028344 which will be used to establish corrective actions that will prevent recurrence on future exams. Along with #33, I made errors while trying to test a fundamental concept in an operationally valid manner. Bank question will be inactivated.
Recommended Changes to the 2023 Wolf Creek NRC Initial Written Exam.
Recommendation:
Remove flawed question from the exam since there is no correct answer for the given conditions. There is no Equipment Malfunction that will require the crew to perform an Emergency Boration per OFN BG-009 with Rods in Manual and steady Reactor Power. It is also difficult to discern which parameter changed the least by either magnitude or percentage from original value when each parameter has a unique unit, and all move in a definite cause-and-effect relationship with each other in the downward direction (Including Reactor Power by Nuclear Instrument indication).
Recommended Changes to the 2023 Wolf Creek NRC Initial Written Exam.
Original Question:
63 ID: 116363 Points: 1.00 Given:
The reactor tripped due to a loss of offsite power.
The crew transitioned to EMG ES-06, NATURAL CIRCULATION COOLDOWN WITH STEAM VOID IN VESSEL (WITH RVLIS), due to steam void formation in the reactor vessel.
While continuing RCS Cooldown and depressurization, RVLIS Natural Circulation range indication dropped below 70%.
Based on these conditions, what action should the crew take per EMG ES-06?
A.
Manually actuate SI B.
Raise Charging flow C.
Re-energize PZR heaters D.
Raise cooldown rate Answer:
C Answer Explanation (Accept accepting both B and C as Correct Answers)
Answer choices sorted. Correct answer at C.
A. Distractor 1 (Manually actuate SI) is INCORRECT, but plausible. EMG ES-06 Fold out page step 1 directs this action if PZR level cannot be maintained > 7% or for subcooling <30°F. This choice is wrong because EMG ES-06 does not provide guidance to actuate SI based on lowering RVLIS Level. This answer choice was selected as the correct answer by a Licensed Operator during validations.
B. Distractor 2 (Raise Charging flow) is INCORRECT, but plausible. EMG ES-06, Step 9 (Continuous Action Step), directs the crew to control charging and letdown as necessary to establish PZR level >20%. This choice is wrong since EMG ES-06 does not provide guidance to raise charging flow in response to lowering RVLIS Level. This answer choice was selected as the correct answer by a Licensed Operator during validations.
C. CORRECT (Re-energize all back-up heaters) EMG ES-06, Step 10 (Continuous Action Step), directs the crew to repressurize the RCS to maintain RVLIS natural circulation range
>70%. Energizing PZR Back-Up heaters will re-pressurize the RCS, and act to drive water from the pressurizer back to the vessel restoring RVLIS Level.
C. Distractor 3 (Raise cooldown rate) is INCORRECT, but plausible. EMG ES-06, Step 8 directs the crew to continue RCS Cooldown and Depressurization while recording RCS and PZR parameters during cooldown using STS BB-011, maintaining RCS Subcooling and maintaining RCS temperature and pressure within limits. The RNO provides direction to control pressure, as necessary, not adjust cooldown rate. This choice is wrong since EMG ES-06 does not provide guidance to raise cooldown rate in response to lowering RVLIS Level.
Recommended Changes to the 2023 Wolf Creek NRC Initial Written Exam.
Question 63 Info Time to Complete:
2 Difficulty:
2.00 System ID:
116363 User-Defined ID:
(23) LO116363
Reference:
NAT CIRC VOIDS Topic:
- 63 - WE10 / EA2.11 (None) - B/CA/3/C RO Importance Rating:
3.5 SRO Importance Rating:
CFR: 41.10 K/A Number:
WE10 EA2.11 Comments:
2023 Wolf Creek NRC Exam, K/A Statement: Natural Circulation with Steam Void in Vessel with/without RVLIS:
Ability to determine and/or interpret the following as they apply to Natural Circulation with Steam Void in Vessel with/without RVLIS: Reactor vessel level.
Tier 1 Group 2 Comprehension / Analysis Technical
References:
LO1732317, EMG ES-04/05/06 NATURAL CIRCULATION COOLDOWN, Objective 11, Explain the basis and any knowledge requirement for EMG ES-06 procedure steps (Page 34).
EMG ES-06, NATURAL CIRCULATION COOLDOWN WITH STEAM VOID IN VESSEL (WITH RVLIS) (Pages 18-20).
Question meets Tier 1 requirements since it tests an applicants knowledge of how to safely operate the plant during emergency and abnormal conditions. This objective is met using: (1) information contained in the sites procedures, including Emergency operating procedures (EMG ES-06), and their associated bases documents, (2) diagnosis that leads to selection of the procedures that should be used to respond to the evolution, (3) the progression of an event; and (4) the assessment of the integrated plant response to emergency or abnormal situations crossing several plant systems or safety functions, or both.
Meets the K/A since the question provides a Natural Circulation with Steam Void in the Vessel with RVLIS scenario and the question tests the required response per EMG ES-06 based on Reactor Vessel Level.
Question History:
Bank Question Rev 0 - Submitted for 2023 Wolf Creek NRC Exam.
Recommended Changes to the 2023 Wolf Creek NRC Initial Written Exam.
New Information that Supports regrade request:
The applicants were briefed per NUREG 1021, ES1.2, paragraph 8 When answering a question, do not make assumptions about conditions that are not specified in the question unless they occur as a consequence of other conditions that are stated in the question The stem does not specify a PZR Level to make the action taken by the crew to raise charging flow wrong (Answer choice B). While the crew is cooling down and depressurizing, they will be utilizing auxiliary spray, which requires the charging pump discharge isolated from the normal charging header for the time that it is aligned to the auxiliary spray header. See Steps 7b and 7c.
EMG ES-06, NATURAL CIRCULATION COOLDOWN WITH STEAM VOID IN VESSEL (WITH RVLIS)
Recommended Changes to the 2023 Wolf Creek NRC Initial Written Exam.
Depending on how rapidly the crew reduces RCS pressure, they might also be losing PZR Level due to contraction of the water from the ongoing cooldown. Therefore, when the crew recognizes that RVLIS indication is less than 70%, they will have to stop depressurizing (realign to the normal charging header) and possibly have to raise charging flow to help counteract the contraction experienced while they were depressurizing. Depending on the PZR Level (not given), the crew may be performing both actions simultaneous to make up inventory during the cooldown if PZR Level is <20%. If the RVLIS indication is dropping as a result of Reactor Head bubble growth and PZR Level were >20% and rising, Answer choice B could be defended as a wrong action. Answer choice C is the most effective way to raise RCS pressure, but with the given information in the stem of the question, Answer Choice B cannot be eliminated as a potential correct answer.
The directed RNO action for EMG ES-06, Step 10 is to Repressurize RCS to maintain RVLIS natural circulation range >70%. Not specifically Reenergize PZR heaters as is directed by Step 9b RNO.
Establishing charging flow will also raise pressure since that action is performed along with stopping the depressurization using Auxiliary Spray.
Recommended Changes to the 2023 Wolf Creek NRC Initial Written Exam.
Exam Author Perspective: This question was selected from the bank and was last used in 2015. I failed to recognize the lack of PZR Level in the stem conditions to eliminate answer choice B as a potential second correct answer challenge. This exam construction error is documented on CR10028344 which will be used to establish corrective actions that will prevent recurrence on future exams. The bank question will be fixed by specifying in the stem that PZR level is 90% and rising which will make the selected correct answer to re-energize PZR Heaters right for two separate reasons (Step 9b RNO and Step 10 RNO) and to make B clearly wrong as well.
Recommendation:
Since PZR Level was not given, accept both B and C as Correct Answers since both actions may be in progress simultaneously, and while establishing charging flow, auxiliary spray is stopped which will stop RCS depressurization and also result in RCS pressure rising.
Internal Use Only Post Exam Lesson Plan Enhancement Suggestions:
Comment ID 19794, IMPACT 2023-7664 - LO4710205, GEN 00-004, NORMAL POWER OPERATIONS (PART 1) Post Exam IMPACT - Enhance Lesson plan to discuss use of SLIM Controller and allow each student the opportunity to practice control of 2nd MFP when paralleling per SYS AE-121. (CR10028341)
Comment ID 19795, IMPACT 2023-7665 - LO1406400, Emergency Diesel Generator System (Mechanical), Objective 5, Explain the function of the starting air subsystem and its controls. (Pages 28-37) Post Exam IMPACT - Enhance Lesson Plan to specify PG19G and PG20G power supplies for Starting Air Compressors.
[Comment ID 17796, IMPACT 2023-7666] - LO1732346, EMG FR-H1, LOSS OF HEAT SINK, Objective 3, Explain the basis and any knowledge requirements for EMG FR-H1 Procedure Steps.
(Page 15) Post Exam IMPACT - Improve Step 44 description for key temperatures and levels for restoration of AFW Feed.
[Comment ID 19797, IMPACT 2023-7667] - LO1408803, Auxiliary, Fuel, and Control Building HVAC Systems, Objective 2, Discuss the function of major Control Building HVAC System components and controls. (Pages 8, 26 and 27) Post Exam IMPACT - Improve description of Inadvertent CRVIS and use of SYS GK-121, CONTROL BUILDING HVAC OPERATION for restoration from CRVIS Lineup.
[Comment ID 17797, IMPACT 2023-7668] - LO4701301, Engineered Safety Features Actuation System (ESFAS), Objective 2, Recognize alarms associated with the Engineered Safety Features Actuation System (Page 19). Post Exam IMPACT - Improve description of Inadvertent CRVIS and use of SYS GK-121 and/or SYS GK-122 as RNO Actions
[Comment ID 19799, IMPACT 2023-7669] - LO1506205, Power Block AC Electrical Distribution, Objective 4, Explain the purpose/function of the Class 1E electrical buses (Page 32). Post Exam Impact - Incorporate data from Change Package 20103 (WIP-E-11023-010-B-1) and ALR 00-022E.
Post Exam Procedure Enhancement Suggestions:
PCR 1032, ALR 00-028C, Step 2b - Fix Typo.
PCR 1033, EMG ES-03, Step 20 - Evaluate use of Ensure instead of Check when establishing Excess Letdown to the VCT. (CR10028338)
WC 2023 NRC Written Exam Analysis no NAMES VALUE:
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 KEY ANSWERS: A D
C B
B D
A D
A C
C A
A B
D C
D C
C D
A B
D D
C B
D D
B A
D B
A B
C C
B D
B C
X B
A A
B C
A D A/D A
NAME TOTAL%
PTS 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 50 93.24 69.0 A
C 82.43 61.0 C
C C
C D
C D
C 89.19 66.0 C
D 78.38 58.0 A
C A
C C
A A
C A
81.08 60.0 C
A C
C D
B D
B 85.86 85.0 C
C A
C C
C 84.85 84.0 A
C A
A D
D B
93.94 93.0 C
D D
C 92.93 92.0 D
80.81 80.0 A
C A
C A
A C
C C
91.92 91.0 B
C B
D A
Total Missed:
3 3
1 4
4 3
2 3
8 1
3 1
2 3
1 3
5 5
1 3
1 1
VALUE:
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
KEY ANSWERS: C A
A B
A D
D B
C D
B A
C C
B C
C D
A B
D B
D C
B B
D C
D A
B C
A B
A D
A C
D B
A D
D C
A C
D C
B B
NAME SRO %
RO %
51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 85 86 87 88 89 90 91 92 93 94 95 96 97 98 99 100 100.0 93.24 C
A B
100.0 82.43 B
B B
C D
100.0 89.19 B
B C
D B
A 100.0 78.38 C
C C
C D
B B
81.08 A
C B
A D
B 80.00 87.84 C
C B
C D
B B
B 84.00 85.14 D
C B
B B
A C
B 92.00 94.59 B
B 92.00 93.24 C
A B
A B
D 76.00 82.43 A
A C
D A
C B
C D
C 92.00 91.89 B
B B
Total Missed:
3 2
8 1
1 3
2 7
3 4
1 2
1 3
2 1
4 1
1 3
1 2
1 1
1 2
2 1
99 74 25 87.2 86.0 86.8 WC 2023 NRC WRITTEN EXAM ANALYSIS 1-75 RO / 76-100 SRO Only 1-75 RO / 75-100 SRO Only SRO Total Points =
RO Only Points =
Key changes: Q33 correct answer changed from D to A, Delete Q41, Q49 accept A or D SRO Only Points =
High miss questions in red: Q29 8/11 missed, Q57 8/11 missed, Q63 7/11 missed RO Questions Average =
GFE questions 70-75 SRO Questions Average =
Class Average =
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