IR 05000482/2023301
| ML23352A071 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 12/19/2023 |
| From: | Heather Gepford NRC/RGN-IV/DORS/OB |
| To: | Reasoner C Wolf Creek |
| References | |
| EPID L-2023-OLL-0042, 50-482/2023-01 50-482/OL-2023 | |
| Download: ML23352A071 (22) | |
Text
December 19, 2023
SUBJECT:
WOLF CREEK GENERATING STATION - NRC EXAMINATION REPORT 05000482/2023301
Dear Mr. Reasoner:
On December 8, 2023, the U.S. Nuclear Regulatory Commission (NRC) completed an initial operator license examination at Wolf Creek Generating Station. The enclosed report documents the examination results and licensing decisions. The preliminary examination results were discussed on October 20, 2023, with you and other members of your staff. A telephonic exit meeting was conducted on December 8, 2023, with Mrs. Michelle Meyer, Operations Training Manager, who was provided the NRC licensing decisions.
The examination included the evaluation of five applicants for reactor operator licenses, three applicants for instant senior reactor operator licenses, and three applicants for upgrade senior reactor operator licenses. The license examiners determined that ten of the eleven applicants satisfied the requirements of 10 CFR Part 55, and the appropriate licenses have been issued.
There were five post-examination comments submitted by your staff. Enclosure 1 contains details of this report and Enclosure 2 summarizes post-examination comment resolution.
No findings were identified during this examination. This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely, Heather J. Gepford, Ph.D., Chief Operations Branch Division of Operating Reactor Safety Docket No. 05000482 License No. NPF-42 Enclosures:
1. Examination Report 05000482/2023301 2. NRC Post-Examination Comment Resolution Electronic distribution via LISTSERV Signed by Gepford, Heather on 12/19/23
ML23352A071 SUNSI Review: ADAMS: Non-Publicly Available Non-Sensitive Keyword:
By: kdc Yes No Publicly Available Sensitive NRR-079 OFFICE RIV/DORS/OB RIV/DORS/OB RIV/DORS/OB RIV/DORS/OB C:OB NAME KClayton COsterholtz SHedger NHernandez HGepford SIGNATURE /RA/ /RA/ /RA/ /RA/ /RA/
DATE 12/12/23 12/18/23 12/18/23 12/18/23 12/19/23
U.S. NUCLEAR REGULATORY COMMISSION Examination Report
Docket Number: 05000482
License Number: NPF-42
Report Number: 05000482/2023301
Enterprise Identifier: L-2023-OLL-0042
Licensee: Wolf Creek Nuclear Operating Corp.
Facility: Wolf Creek Generating Station
Location: Burlington, Kansas
Examination Dates: October 16, 2023, to December 8, 2023
Examiners: K. Clayton, Senior Operations Engineer (Chief Examiner)
S. Hedger, Senior Emergency Preparedness Inspector C. Osterholtz, Senior Operations Engineer N. Hernandez, Senior Operations Engineer
Approved By: Heather J. Gepford, Ph.D., Chief Operations Branch Division of Operating Reactor Safety
Enclosure 1 SUMMARY
Examination Report 05000482/2023301; October 16 to December 8, 2023; Wolf Creek Generating Station; Initial Operator Licensing Examination Report
The NRC examiners evaluated the competency of five applicants for reactor operator licenses, three applicants for instant senior reactor operator licenses, and three applicants for upgrade senior reactor operator licenses at Wolf Creek Generating Station.
The licensee developed the examinations using NUREG-1021, "Operator Licensing Examination Standards for Power Reactors," Revision 12. The written examination was administered by the licensee on November 1, 2023. The NRC examiners administered the operating tests from October 16-20, 2023.
The NRC examiners determined that ten of the eleven applicants satisfied the requirements of 10 CFR Part 55, and the appropriate licenses have been issued.
A. NRC-Identified and Self-Revealing Findings
None.
B. Licensee-Identified Violations
None.
REPORT DETAILS
OTHER ACTIVITIES - INITIAL LICENSE EXAM
.1 License Applications
a. Scope
The NRC examiners reviewed all license applications submitted to ensure each applicant satisfied relevant license eligibility requirements. The NRC examiners also audited three of the license applications in detail to confirm that they accurately reflected the subject applicants qualifications. This audit focused on the applicants experience and on-the-job training, including control manipulations that provided significant reactivity changes.
b. Findings
No findings were identified.
.2 Examination Development
a. Scope
The NRC examiners reviewed integrated examination outlines and draft examinations submitted by the licensee against the requirements of NUREG-1021. The NRC examiners conducted an onsite validation of the operating tests.
b. Findings
The NRC examiners provided outline, draft examination, and post-validation comments to the licensee. The licensee satisfactorily completed comment resolution prior to examination administration.
Although the written examination was determined to be within the acceptable quality range, five questions were challenged during the post examination review. In accordance with NUREG-1021, future written examination submittals need to incorporate lessons learned.
.3 Operator Knowledge and Performance
a. Scope
On November 1, 2023, the licensee proctored the administration of the written examinations to all eleven applicants. The licensee staff graded the written examinations, analyzed the results, and presented their analysis and post-examination comments to the NRC on November 16, 2023.
The NRC examination team administered the various portions of the operating tests to all applicants from October 16-20, 2023.
b. Findings
No findings were identified.
Ten of the eleven applicants passed the written examination, and all eleven applicants passed the operating test. The final examinations and post-examination analysis and comments may be accessed in the ADAMS system under the accession numbers noted in the attachment.
The examination team noted one generic weakness associated with applicant performance on the operating test. The applicants struggled with procedure use and adherence with action verbs ensure versus verify. Post-examination analysis revealed three generic weaknesses associated with applicant performance on the written examination. These weaknesses were captured in the licensees corrective action program as Condition Reports 2023-10028338 and 2023-10028344. Copies of all individual examination reports were sent to the facility Training Manager for evaluation and determination of appropriate remedial training.
.4 Simulation Facility Performance
a. Scope
The NRC examiners observed simulator performance with regard to plant fidelity during examination validation and administration.
b. Findings
No findings were identified.
.5 Examination Security
a. Scope
The NRC examiners reviewed examination security for examination development during both the onsite preparation week and examination administration week for compliance with 10 CFR 55.49 and NUREG-1021. Plans for simulator security and applicant control were reviewed and discussed with licensee personnel.
Findings
No findings were identified.
EXIT MEETINGS AND DEBRIEFS
Exit Meeting Summary
The chief examiner presented the preliminary examination results to Mr. J. McCoy, Site Vice President, and other members of the staff on October 19, 2023. A telephonic exit was conducted on December 8, 2023, between Mr. K. Clayton, chief examiner, and Mrs. Michelle Meyer, Operations Training Manager. The licensee provided proprietary information in the post examination analysis; these materials have been returned.
ADAMS DOCUMENTS REFERENCED
Accession No. ML23346A105 - FINAL WRITTEN EXAMS Accession No. ML23346A106 - FINAL OPERATING TEST Accession No. ML23346A099 - POST-EXAMINATION ANALYSIS-COMMENTS
NRC Resolution to Wolf Creek Post-Examination Comments
A complete text of the licensee's post-examination analysis and comments can be found in ADAMS under Accession Number ML23346A099.
In accordance with ES4.4 of NUREG-1021, page 7:
If the facility licensee recommends deleting or changing the answers to four or more of the questions on an RO written examination (or two or more on an SRO only exam) that it developed, consider asking the facility licensee to explain why so many post-examination changes were necessary and what actions will be taken to improve future license examinations. As discussed in ES-5.1, Issuing Operator Licenses and Post-Examination Activities, the NRC will also consider post examination deletions and changes when evaluating the quality of the facility licensees proposed examination for documentation in the examination report.
Wolf Creek has written Condition Report 2023-10028344 to address why so many changes were necessary and what actions are being taken to improve future examinations.
Questions 33, 41, 49, 57, and 63 had post exam comments with requested changes and appear below in numerical order. It is also noted that none of the applicants asked for clarification on any of the challenged questions during the written exam administration on November 1, 2023.
RO QUESTION # 33
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:
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.
Enclosure 2 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).
The facility licensee and an applicant contend that the keyed answer LICENSEE COMMENT:
D: indicate the calibrated reference junction temperature, is incorrect for Wolf Creek Generating Station. They provided new information for this question with plant data for both failed open thermocouples (that read 0°F) as well as data captured during reactor vessel head removal. During this evolution, thermocouples are disconnected and reading 0°F in their system (indicative of an open circuit). Finally, the facility provided a schematic that shows that an open circuit thermocouple in their system is connected in series not parallel and therefore an open circuit condition on the detector will display as 0°F.
The facility exam writer started with a bank generic fundamentals exam (GFE) question to test knowledge of a failed open thermocouple (the bank question is listed below):
Topic: 191002 Knowledge: K1.14 QID: P213
An open circuit in a thermocouple detector causes the affected temperature indicator to fail
A. high B. low C. reference junction temperature D. as-is
Answer: C reference junction temperature
The exam writer created a site-specific question on this bank question. However, during the post-exam review on November 2, 2023, an applicant who worked in reactor engineering (before starting initial licensing class) brought up new information that a thermocouple was failed open in the plant, and it was not reading the reference junction temperature but
2 degrees F. Because of this newly discovered technical information, the facility licensee recommended changing the key from D to A.
NRC RESOLUTION: The NRC agrees with the licensees recommendation. Based on a review of the plant data for this question, the NRC agrees that A, indicate 0°F is the correct answer, and that D, indicate the calibrated reference junction temperature is incorrect. The NRC chief examiner reviewed the industry training materials Sensors and Detectors Part 1, Revision 4, and believes the source of the original GFE bank error was a knowledge check question within the training materials. It has an incorrectly keyed answer which is apparent when this original knowledge check question is reviewed. It is listed below:
In this question the checkmark for the correct answer is C but if the C explanation is read it explains that the temperature indication will fail low. Therefore, the keyed answer for this knowledge check question should have been B low. While changing the original GFE bank question into a site-specific question, no one on the licensees staff verified that their actual system design was consistent with the GFE bank question.
The NRC agrees with the licensees recommendation that A is the correct answer for question 33 and the key has been edited to reflect this change.
3 RO QUESTION # 41
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:
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.
The licensee stated in their post-exam comments that: LICENSEE COMMENT:
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.
During the post examination review with the class on November 2, 2023, an applicant provided feedback of this flaw by stating on his feedback form Remove question from the exam due to conditions given in stem being not plausible. Steam Generators on a Large Break LOCA are not coupled to the RCS. More information is attached. In his attachment, he quoted the bases for step 1 (paraphrased) for Emergency Procedure EMG FR-H1:
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.
The licensee also stated how the question became flawed from the original bank question in their post examination comments: 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.
Due to the flawed stem from this unintended change, the licensee recommended deleting this question from the exam.
NRC RESOLUTION: The NRC agrees with the licensees recommendation. The unintended insertion of LB LOCA into the stem of the question created a question that is not plausible for the given conditions and the possible choices. This question has been deleted from the exam and the key has been edited to reflect this change.
5 RO QUESTION # 49
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:
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
6 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.
LICENSEE COMMENT: In the post examination comments, the licensee states that:
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.
During the post-exam review on November 2, 2023, an applicant also challenged (on a question feedback form) that there are two correct answers for the given stem based on lack of a mode statement in the stem:
Question (49) has 2 correct answers. Answers A and D are both correct in Mode 5 and would be up to SRO discretion based on plant conditions not given in the stem. Fix bank question to specify offsite power was restored in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
NRC RESOLUTION: The NRC agrees with the licensees recommendation. Based on a review of the plant data provided for this question and the comments during post-exam review from an applicant, the NRC agrees that A and D are both correct when no mode or timeframe (from full power to the loss of offsite power) are provided in the stem. The key has been edited to reflect this change.
7 RO QUESTION # 57
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:
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.
8 The licensee recommended deleting the question based on new LICENSEE COMMENT:
information provided as a result of running the exam question setup on the full scope plant-referenced simulator. Specifically, the licensee concluded there was no correct answer for the given question. First, they determined was difficult to discern which parameter changed the least by either magnitude or percentage from original value. Additionally, they contended that because 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, the information in the stem confused the applicants.
In the simulator, with main turbine controls in first stage pressure mode, the main control valves opened slightly to maintain a constant first stage pressure. All four parameters lowered: The boration lowered Tavg, which also lowered program PZR level, and resulted in lowering S/G pressure. The reactor power indication by Nuclear Instruments 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.
Comparison of Key Parameters:
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%
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. 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.
The licensee post-examination comments also stated that the 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 the results might be impacted by that arbitrary equipment malfunction statement.
NRC RESOLUTION: The NRC disagrees with the licensees recommendation to delete the question. This question originated as a bank question on reactor fundamentals used on the Callaway 2002 Initial NRC Exam.
The NRC acknowledges that newly discovered technical information supporting a change in the answer key, if identified and adequately justified by the facility licensee, may warrant examination changes (NUREG-1021, ES-4.4, section C.3.c). However, the information provided did not conclusively demonstrate that the originally keyed answer was incorrect.
Based on the simulator data, one could conclude that auctioneered high Tavg returns closest to its original value compared to the other three parameters, if you assume the nuclear instruments accurately reflect initial and final reactor power. However, the simulator data showed that core thermal power returned to within 0.1% of its original power, thus being closer to initial than Tavg.
It is noted that the licensees response also stated that:
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. [Emphasis added]
Thus the contention that there is no correct answer because it was difficult to discern which parameter changed the least has not been conclusively demonstrated by the simulator data since the final reactor power value (prior to recalibration of the NIs) provided in the table is not the most accurate indication of reactor power, a detail the applicants should have known.
Regardless of the potential inaccuracy associated with the final reactor power value, the NRC acknowledges that the simulator data, indicative of todays control systems (compared to 2002),
seems to show that this question is measuring minutia if you were attempting a quantitative assessment and answer. Notwithstanding, NUREG-1021, ES-4.4, section C.3.d states that the NRC will not accept examination changes for a question that tests minutiae, even though the facility licensee and the NRC previously agreed that the question did not test minutiae.
The question was intended to test fundamental knowledge, and the licensees own training documents support that the applicants should have been able to answer the question without knowing specific values for the parameters. Both Lesson Plan LO1732419, OFN BG-009, Emergency Boration, and Lesson Plan LO1130641, Reactor Theory - Reactor Operational Physics discuss the impact of boration on reactor power, pressurizer pressure, RCS temperature, etc.
Lesson Plan LO1732419 included the Note found in procedure OFN BG-009: Prolonged Emergency Boration with the Plant at power may require the Plant to be tripped due to rapid RCS Temperature, Pressure, and PZR Level decreases.
Lesson Plan LO1130641 included three learning objectives related to this question:
Objective 19. Explain the effects of control rod motion or boration/dilution on reactor power.
Objective 21. Explain the relationship between steam flow and reactor power given specific conditions. [i.e. Ovation in first stage pressure mode]
Objective 22. Explain reactor response to a control rod insertion. [i.e. reactivity change]
10 The lesson plan included summary questions covering the concepts. For example (emphasis added):
Q: For a reactor operating at 60 percent power, describe the effect that a boron dilution will have on power and RCS average temperature. (Assume rod position and turbine load remain constant.)
A: Power will increase, resulting in an increase in RCS average temperature. This, in turn, will cause reactor power to decrease to approximately 60 percent power.
Q. A reactor is operating in steady state at 80% power. What effect will a boron concentration dilution from 450 ppm to 430 ppm have on the reactor with control rods in manual?
A. The boron dilution gradually adds positive reactivity. In response, power increases slightly, causing moderator temperature to increase. With control rods in manual, the net effect will be power essentially unchanged at 90%, but coolant temperature will be higher.
Q: A reactor is operating at 50% power when the operator inserts control rods a short distance.
Describe how reactor power will respond, and how will the new steady state power level compare to the initial power level?
A: The rods insert negative reactivity, causing a decrease in reactor power. With steam demand constant, the mismatch between power and steam demand causes Tave to decrease. The temperature decrease inserts positive reactivity, bringing power back up to meet steam demand.
Thus, the fundamental reactor concept of how reactivity (and/or steam demand) changes impact temperature, reactor power, etc. were covered during initial licensed operator training.
Additionally, the cited examples indicate that the concepts were discussed qualitatively rather than quantitatively as the simulator data implied. The lesson plans instructed reactor power was essentially unchanged or, using the terminology of the exam returned closest to its pre-boration value, following reactivity changes. It is noted that this information further supports the NRC conclusion that minutia was not being tested by the question.
The second aspect of the licensees contention was that the stem focus was unclear in that the equipment malfunction that required an emergency boration confused the students. Concerns with the stem are addressed in NUREG-1021, section C.3.c of ES-4.4: The NRC will consider examination changes for the following types of errors, if identified and adequately justified by the facility licensee or an applicant:
- a question with an unclear stem that confused the applicants or did not provide all the necessary information (to assist in determining whether an unclear stem confused the applicants, closely evaluate any applicant questions asked during the examination; also evaluate the question stem to determine whether the information provided could reasonably result in the applicant misunderstanding the intent of the question or the validity of the answer choices)
The NRC noted that no questions were asked about Question 57 by any applicant during exam administration. The NRC also evaluated the question stem to determine whether the information provided could reasonably result in an applicant misunderstanding the intent and concluded that the stem was not focused on the cause of the emergency boration or the equipment malfunction that required the boration but rather the plant response with rods in manual mode and the turbine in first stage pressure mode. Additionally, any speculation about what equipment malfunction drove the need for emergency boration would have been making
assumptions about conditions not specified in the stem, contrary to the NUREG-1021, ES1.2, briefing each applicant received.
Given the above, the NRC disagrees with deleting the question and no change to the key is required.
RO QUESTION # 63
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:
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.
12 D. 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.
The licensee recommended accepting both B and C as correct LICENSEE COMMENT:
answers because pressurizer level was not provided in the stem of the question.
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 7 of emergency procedure EMG ES-06, NATURAL CIRCULATION COOLDOWN WITH STEAM VOID IN VESSEL (WITH RVLIS).
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.
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.
During the post-exam review on November 2, 2023, an applicant also challenged (on a question feedback form) that there were two correct answers for the given stem (answers B and D)
because pressurizer level was not given in the stem:
13 Without giving pressurizer level in the stem it is also possible to repressurize the reactor coolant system by raising charging flow. Accept B as a second correct answer.
NRC RESOLUTION: The NRC disagrees with the licensees recommendation. For distractor B to be correct, the applicant would have to assume that pressurizer level was in an abnormal condition requiring action. This is contradictory to the guidance regarding assumptions in NUREG-1021, section ES-1.2, which was quoted in part by the licensee in their response.
Paragraph 8 continues by providing an example: [Y]ou should not assume that any alarm has activated unless the question so states Therefore, as each of the applicants was briefed on this information, they should not have been making assumptions about potential abnormal indications that were not stated in the stem. The NRC also noted that none of the applicants asked a clarifying question during administration on what pressurizer level was when RVLIS dropped below 70%. So although pressurizer pressure was not stated in the stem, and Step 9 of procedure EMG ES-06 to control charging/letdown to establish pressurizer pressure > 20% is a continuous action step, the applicant would have to assume the pressure was in an abnormal condition for distractor B Raise Charging flow to be a possible correct answer.
The NRC also noted that the question stem directly asked for what actions should be taken when RVLIS indication was abnormal, not what actions should be taken when pressurizer pressure was abnormal. The stem of the question was focused on the response not obtained column of emergency procedure EMG ES-06 when RVLS drops below 70%. The step of the procedure is captured below:
The purpose of this procedure is to monitor void growth when there is a void in the upper reactor vessel head region (as indicated in the procedure title) and specifically requires RVLS to stay above 70% so the void or bubble can continue to be monitored as the reactor is cooled down and depressurized. At this point in the procedure, charging is in manual control to prevent pressurizer level from going too high and going solid; as such, repressurizing the RCS to maintain RVLS greater than 70% is accomplished by re-energizing pressurizer heaters while charging remains in manual mode.
Another salient point on this 2015 bank question was that it was taken directly from Wolf Creeks exam bank by the exam author with no performance issues noted. The explanation in this question (ID 116363) on the basis for why increase charging flow was incorrect is a fundamental concept for a Westinghouse pressured water reactor design is that it states that Increasing charging flow will not necessarily directly re-pressurize the RCS. The bubble in the head could maintain, and the water would go into the pressurizer, not refilling the core until the pressurizer is mostly solid.
Given the above discussion, the NRC concluded that the only correct answer for this question is C Re-energize PZR heaters. No change to the key is required for this question.
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