IR 05000266/2021301
| ML21252A136 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 09/14/2021 |
| From: | Patricia Pelke NRC/RGN-III/DRS/OLB |
| To: | Strope M Point Beach |
| Zoia C | |
| Shared Package | |
| ML20139A043 | List: |
| References | |
| Download: ML21252A136 (16) | |
Text
September 14, 2021
SUBJECT:
POINT BEACH NUCLEAR PLANT - NRC INITIAL LICENSE EXAMINATION REPORT 05000266/2021301; 05000301/2021301, AND 07200005/2021301
Dear Mr. Strope:
On August 12, 2021, the U.S. Nuclear Regulatory Commission (NRC) completed the initial operator licensing examination process for license applicants employed your Point Beach Nuclear Plant. The enclosed report documents the results of those examinations. Preliminary observations noted during the examination process were discussed on August 2, 2021, with Mr. M. Holzmann and other members of your staff. An exit meeting was conducted by telephone on August 25, 2021, between you and Mr. C. Zoia, Senior Operator Licensing Examiner, to review the proposed final grading of the written examination for the license applicants. During the telephone conversation, NRC resolutions of the facilitys post-examination comments, initially received by the NRC on August 12, 2021, were discussed.
The NRC examiners administered an initial license examination operating test during the week of July 26, 2021. The written examination was administered by Point Beach Nuclear Plant training department personnel on August 3, 2021. Five Senior Reactor Operator (SRO) and four Reactor Operator (RO) applicants were administered license examinations. The results of the examinations were finalized on August 24, 2021. Eight applicants passed all sections of their respective examinations. One applicant was issued a Preliminary Results letter. Five applicants were issued senior operator licenses and three applicants were issued operator licenses.
The as-administered written examination and operating test, as well as documents related to the development and review (outlines, review comments and resolution, etc.) of the examination will be withheld from public disclosure until August 12, 2023. However, because one applicant received a preliminary results letter due to receiving a non-passing grade on the written examination, the applicant was provided a copy of the RO written examination. For examination security purposes, your staff should consider this written examination material uncontrolled and exposed to the public. 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 Title 10 of the Code of Federal Regulations, Part 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely, Signed by Pelke, Patricia on 09/14/21 Patricia J. Pelke, Chief Operations Branch Division of Reactor Safety Docket Nos. 50-266, 50-301, and 72-005 License No. DPR-24 and DPR-27 Enclosure:
1. Examination Report 05000266/2021301; 05000301/2021301; and 07200005/2021301 2. Post-Examination Comments, Evaluation, and Resolutions 3. Simulator Facility Fidelity Report cc: Distribution via LISTSERV M. Hastings, Training Manager
SUMMARY
ER 05000266/2021301; 05000301/2021301; 07200005/2021301; 07/26/2021-08/12/2021;
NextEra Energy Point Beach, LLC, Point Beach Nuclear Plant; Initial License Examination Report.
The announced initial operator licensing examination was conducted by regional Nuclear Regulatory Commission (NRC) examiners in accordance with the guidance of NUREG-1021,
Operator Licensing Examination Standards for Power Reactors, Revision 11.
Examination Summary:
Eight of nine applicants passed all sections of their respective examinations. Five applicants were issued senior operator licenses and three applicants were issued operator licenses. One applicant was issued a Preliminary Results letter for failure of one section of the administered examination. (Section 4OA5.1).
REPORT DETAILS
4OA5 Other Activities
.1 Initial Licensing Examinations
a. Examination Scope
The NRC examiners and members of the facility licensees staff used the guidance prescribed in NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 11, to develop, validate, administer, and grade the written examination and operating test. The written examination outlines were prepared by the NRC staff and were transmitted to the facility licensees staff. Members of the facility licensees staff developed the operating test outlines and developed the written examination and operating test. The NRC examiners validated the proposed examination during the week of June 21, 2021, with the assistance of members of the facility licensees staff. During the on-site validation week, the examiners audited two license applications for accuracy. The NRC examiners, with the assistance of members of the facility licensees staff, administered the operating test, consisting of job performance measures and dynamic simulator scenarios, during the period of July 26, 2021, through July 30, 2021. The facility licensee administered the written examination on August 3, 2021.
b. Findings
- (1) Written Examination The NRC examiners determined that the written examination, as originally proposed by the licensee, was within the range of acceptability expected for a proposed examination.
Less than 20 percent of the proposed examination questions were determined to be unsatisfactory and required modification or replacement.
During validation of the written examination, several questions were modified or replaced. All changes made to the proposed written examination, were made in accordance with NUREG-1021, Operator Licensing Examination Standards for Power Reactors, and documented on Form ES-401-9, Written Examination Review Worksheet. The Form ES-401-9, the written examination outlines, and both the proposed and final written examinations, will be available electronically in the NRC Public Document Room or from the Publicly Available Records component of NRCs Agencywide Documents Access and Management System (ADAMS) on August 12, 2023, (ADAMS Accession Numbers ML20139A042, ML20139A040, ML20139A041, and ML20139A046 respectively).
On August 12, 2021, the licensee submitted documentation noting that there were two post-examination comments for consideration by the NRC examiners when grading the written examination. The post-examination comments and the NRC resolution for the post-examination comments, are provided in Enclosure 2 to this report.
The NRC examiners graded the written examination on August 16, 2021, and conducted a review of each missed question to determine the accuracy and validity of the examination questions.
- (2) Operating Test The NRC examiners determined that the operating test, as originally proposed by the licensee, was within the range of acceptability expected for a proposed examination.
Less than 20% of the proposed operating test portion of the examination was determined to be unsatisfactory and required modification or replacement.
During the validation of the operating test, several Job Performance Measures (JPMs)were modified or replaced, and some modifications were made to the dynamic simulator scenarios. Changes made to the operating test portion of the examination, were made in accordance with NUREG-1021, Operator Licensing Examination Standards for Power Reactors, and documented on Form ES-301-7, Operating Test Review Worksheet.
The Form ES-301-7, the operating test outlines, and both the proposed and final as administered dynamic simulator scenarios and JPMs, will be available electronically in the NRC Public Document Room or from the Publicly Available Records component of NRC's ADAMS on August 12, 2023, (ADAMS Accession Numbers ML20139A042, ML20139A040, ML20139A041, and ML20139A046 respectively).
The NRC examiners completed operating test grading on August 24, 2021.
- (3) Examination Results Five applicants at the Senior Reactor Operator (SRO) level and four applicants at the Reactor Operator (RO) level were administered written examinations and operating tests. Eight applicants passed all portions of their examinations and were issued their respective operating licenses on August 24, 2021. One applicant was issued a Preliminary Results letter for failure of the written section of the administered examination.
.2 Examination Security
a. Scope
The NRC examiners reviewed and observed the licensee's implementation of examination security requirements during the examination validation and administration to assure compliance with Title10 of the Code of Federal Regulations, Part 55.49, Integrity of Examinations and Tests. The examiners used the guidelines provided in NUREG 1021, Operator Licensing Examination Standards for Power Reactors, to determine acceptability of the licensees examination security activities.
b. Findings
An NRC Exam Security Event occurred on July 29, 2021, during administration of an in-plant JPM. The licensees exam team lost control of the exam material as it was being transferred at the Radiation Protection (RP) checkpoint. Once identified, RP Technicians controlled this material. Approximately 2 minutes elapsed where the material was left in the Small Article Monitor (SAM) without positive control. It was confirmed that no other individuals accessed the SAM prior to RP controlling the exam material.
During this 2 minute period, the applicants were under full sequestration and did not have access to either the exam material or to the individuals that found the improperly controlled material. The two RP Technicians and one RP Manager who had access to the material were signed onto a limited exam security agreement and were briefed on their responsibilities.
The event was promptly discussed with the NRC and it was determined that the applicants did not have access to the material. Therefore, the JPM was not considered compromised and could be administered for the remainder of the day, as planned. This event was documented in Action Request (A/R) number 02399829 and formally investigated per procedure PI-AA-100-1007 Issue Investigation.
4OA6 Management Meetings
.1 Debrief
The chief examiner presented the examination team's preliminary observations and findings on August 2, 2021, to Mr. M. Holzmann, Operations Director, and other members of the Point Beach Nuclear Plant Operations and Training Department staff.
.2 Exit Meeting
The chief examiner conducted an exit meeting on August 25, 2021, with Mr. M. Strope, Site Vice President, by telephone. The NRCs final disposition of the stations grading of the written examination and post-examination comments were disclosed and discussed during the telephone discussion. The chief examiner asked the licensee whether any of the retained submitted material used to develop or administer the examination should be considered proprietary. No proprietary or sensitive information was identified during either the examination or debrief/exit meetings.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- M. Strope, Site Vice President
- M. Holzmann, Operations Director
- B. Woyak, Safety Assurance and Learning Director
- M. Hastings, Training Manager
- C. Hill, Fleet Training Manager
- E. Schultz, Licensing Manager
- B. Vander Velde, Assistant Operations Manager - Training
- J. Bryant, Assistant Operations Manager - Line
- A. Moore, Operations Training Supervisor
- E. Salzwedel, Training Supervisor
- J. Krear, Exam Project Operations Supervisor (EPOS)
U.S. Nuclear Regulatory Commission
- C. Zoia, Senior Operations Engineer, Chief Examiner
- G. Callaway, Senior Operations Engineer, Examiner
- J. DeMarshall, Senior Operations Engineer, Examiner
- J. Robbins, Operations Engineer, Examiner
- T. Hartman, Point Beach Senior Resident Inspector
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened, Closed, and Discussed
None
LIST OF ACRONYMS USED
ADAMS Agency-Wide Document Access and Management System
CI Containment Isolation
NRC U.S. Nuclear Regulatory Commission
RNO Response Not Obtained
RO Reactor Operator
RP Radiation Protection
SAM Small Article Monitor
SRO Senior Reactor Operator
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Question 23
Given the following:
Both units are at Rated Thermal Power
The crew transitions to AOP-40A, Control Room Abandonment Due to Fire
Both units have been tripped
Which of the following describes:
Prior to leaving the control room, the crew will __(1)___ for both units. This is
performed to __(2)___
(CV-200A/B/C, LTDN Orifice A/B/C Outlet CV)
(RC-430, PZR PORV)
(RC-431C, PZR PORV)
A. (1) ensure CV-200A/B/C are shut
(2) prevent spurious operation
B. (1) isolate RC-430 and RC-431C
(2) prevent spurious operation
C. (1) ensure CV-200A/B/C are shut
(2) allow for remote/local operation of these valves
D. (1) isolate RC-430 and RC-431C
(2) allow for remote/local operation of these valves
Answer: B
Applicant Comment (55-75630):
This question has two correct answers.
The stem states prior to leaving the control room, this occurs at step 13 of AOP-40A. The stem
does not state that the CR is NOT habitable, therefore no need to evacuate prior to step 13
exists. The stem does not have enough information to inform the applicant where in the
procedure the question conditions are (i.e., prior to step 13). Therefore, the applicant has no
reason to assume there is a need for evacuation prior to step 13.
AOP-40A, step 8 states to Initiate Containment Isolation for both units. The background for
step 8, Initiate Containment Isolation, purpose states: This step isolates the containments of
both units to prevent possible inadvertent valve operations from adversely affecting safe
shutdown conditions.
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Step 4 states Shut PORV Block Valves vice Isolate RC-430 and RC-431C as given in
answer (B). This is the same action, but different verbiage. Distractor (A) states to ensure CV-
200A/B/C are shut to prevent spurious operation. Initiating Containment Isolation will shut CV-
371 and CV-371A, CVCS LD ISO VLVs. An interlock feature of either CV-371 or CV-371A
going shut, will close CV-200A/B/
AOP-40A. Neither (A) or (B) is worded directly from the procedure and both actions are
performed in AOP-40A, therefore, both (A) and (B) should be considered correct.
Facility Position on Applicant Comment:
The valves in choice A will in fact be closed by operator actions taken during the performance
of AOP-40A, and this action is performed with the same purpose as the reason for ensuring the
actions in choice B PORV block valves are shut, to prevent inadvertent operations of the
valves. The facility licensee concurs with the applicants comment.
NRC Evaluation/Resolution:
Question 23 (Q23) provided the conditions that Both units have been tripped and The crew
transitions to AOP-40A, Control Room Abandonment Due to Fire, and asked:
Which of the following describes:
Prior to leaving the control room, the crew will __(1)___ for both units. This is
performed to __(2)___.
This information clearly indicated that the control room was NOT HABITABLE As result of Fire,
for the following reasons:
The crew would only perform AOP-40A, Steps 2 and 3, to trip both reactors if
the control room was uninhabitable per Step 1, Check Control Room - NOT
HABITABLE As result of Fire. AOP-40A would be exited otherwise.
Leaving the control room per Step 13, Evacuate Control Room, would not be
performed per AOP-40A if the control room was, in fact, habitable, because the
procedure would be exited per Step 1, RESPONSE NOT OBTAINED (RNO),
Do NOT continue unless the Control Room is uninhabitable.
No guidance prior to Step 13 allowed for an early evacuation, nor does any information provided
in Q23 imply a need to leave the control room earlier. Also, Q23 contained sufficient information
to limit its scope between Step 4, Shut PORV Block Valves, and Step 13, Evacuate Control
Room, due to the condition Both units have been tripped.
The applicant correctly stated that Shut PORV Block Valves was the same action as isolate
RC-430 and RC-431C, per distractor B, which was the correct answer given in the approved
answer key.
The applicant also correctly stated that AOP-40A, Step 8, directed a containment isolation (CI)
and that CV-200A/B/C closed indirectly, due to an interlock, when initiating a C
- I. However, the
applicant then incorrectly concluded that distractor A, ensure CV-200A/B/C are shut, was a
second correct answer on this basis. That conclusion was flawed because distractor A said,
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
ensure CV-200A/B/C are shut, but not, shut CV-200A/B/C. The two statements mean very
different things, as further explained below.
Per the licensees procedure OM 3.7, AOP AND EOP PROCEDURE USAGE FOR RESPONSE
TO PLANT TRANSIENTS:
The word Ensure is used to confirm that an expected desirable condition exists.
If the condition does NOT exist, the appropriate contingency, either stated or
implied, is to establish the expected condition.
Also, an example of the meaning of ensure was found in AOP-40A, Step C8 RNO:
IF 1P-29, TDAFW Pump tripped on overspeed, THEN reset as follows:
1. Align PUSH trip arm tab under the flat side of the trip button washer.
2. Move the plug/stem latching mechanism into line by moving the trip rod
to the extreme south position.
3. Ensure the overspeed mechanism is latched.
Steps 1 and 2 in this example, should result in latching the overspeed mechanism. However,
Step 3 added Ensure that the overspeed mechanism is latched. Therefore, the word ensure
goes beyond actions that should result in a desired condition, it requires confirming that result,
or establishing the expected condition if necessary.
For another example, many valves would close as a direct response to a CI, but specific
verification steps were not included in AOP-40A prior to Step 13. This was because with the
control room NOT HABITABLE As result of Fire, no time would exist to safely ensure that all
CI valves are closed, much less to safely ensure that an indirect effect of a CI signal occurred.
Whether or not valves CV-200/A/B/C close was not relevant to Q23. Distractor A was incorrect
because AOP-40A does not direct the crew to ensure that valves CV-200A/B/C were closed.
Specifically, ensure in Q23, distractor A meant, by the given definition, observe CV-200A/B/C
position indication, or close the valves if they are not already closed. The stem of Q23 asked
what had to be done prior to leaving the control room. Ensuring valve positions was not
required in the steps of AOP-40A in the scope this question. As stated earlier, the crew only
proceeded past step 1 and tripped both reactors if the control room was NOT HABITABLE As
result of Fire, due to heat, smoke, or fire in separate control boards. The steps prior to
evacuation of the control room would be limited to only those actions that can be performed
quickly in support placing the plant in a safe and stable condition.
Therefore, the
annotated on the answer key, was the only correct answer, and the question was considered
acceptable as administered. This challenge was denied.
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Question 32
Given the following:
Unit 1 has experienced a Loss of Coolant Accident from Rated Thermal Power
The crew is implementing EOP-1.2, Post LOCA Cooldown and Depressurization
Maximum charging is established
1P-1B, Reactor Coolant Pump is running
Both SI pumps are running
Both RHR pumps are stopped and in AUTO
RCS Hot Leg temperature is 300°F
When evaluating conditions to stop the first SI pump, there is less subcooling than required.
In accordance with the RNO step in EOP-1.2, the operator starts an RHR pump and then
stops the SI pump.
What is accomplished by starting an RHR pump prior to stopping the SI pump?
A. Ensures the RCS will remain subcooled
B. Prevents void formation in the reactor vessel head
C. Prevents a challenge to the Core Cooling critical safety function
D. Ensures conditions are maintained for continued RCP operation
Answer: A
Applicant Comment (55-75630):
This question has two correct answers.
The question stem is unclear, in that it does not provide all necessary information. The stem
does not give a value for subcooling, nor does it give parameters that could be used to calculate
a subcooling value. The only information about subcooling is that there is less subcooling than
required. The issue with this statement is an applicant is not given enough information to
determine which subcooling requirement is not met.
Therefore, the wording of the question is vague enough to have the applicant consider all
subcooling requirements that are currently applicable, including the foldout page subcooling
requirements.
Applicant understood the question stem to mean that less subcooling than required meant
Foldout Page criteria was met for SI Reinitiation Criteria. SI Reinitiation Criteria: If either
condition below occurs, then manually start ECCS pumps as necessary to restore RCS
subcooling and PZR Level: - RCS subcooling based on core exit thermocouples less than [62F]
37F - Pzr Level - cannot be maintained greater than [26%] 11%.
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
From the ERG Background document, ECCS pumps should be restarted in order to reestablish
RCS subcooling. Applicant considered that SI flow was inadequate and that RHR flow should
be initiated based in conditions given.
SI Reinitiation Criteria is based on maintaining core cooling. Therefore, distractor (C) should be
considered correct also.
Facility Position on Applicant Comment:
The fact this was a high miss question is evidence that the flaws as described above were seen
by the majority of applicants. Given the manner in which the term subcooling was used in the
question, the candidates were not able to determine if the foldout page criteria was met or not.
Also, a similar style question, where procedure step RNOs were used as distractors and the
correct answer was based on foldout page actions was used on this exam. The facility licensee
concurs with the applicants comment.
NRC Evaluation/Resolution:
The applicant contended that not enough information was given to calculate a subcooling value
or determine which subcooling requirement was not met. The applicant then considered all
subcooling requirements that were applicable and concluded that Question 32 (Q32) meant SI
REINITIATION CRITERIA on the FOLDOUT PAGE FOR EOP-1.2 UNIT 1, which provided a
basis for a second correct answer. However, Q32 asked for what was accomplished by a
RESPONSE NOT OBTAINED (RNO) step from the body of procedure EOP-1.2, when
evaluating conditions to stop the first SI pump, given that there was less subcooling than
required. No subcooling value or calculation was requested, and no clarifying information was
sought about Q32 by any applicant during exam administration.
The applicant stated that distractor A, Ensures the RCS will remain subcooled, was one of
two correct answers, but that distractor (C) should be considered correct also. Distractor A
was the correct answer per the approved key, and was supported by POINT BEACH NUCLEAR
PLANT EOP BACKGROUND DOCUMENT, BG EOP-1.2, POST LOCA COOLDOWN AND
DEPRESSURIZATION, for Step 11, which stated, in part:
If the RCS subcooling criterion is not satisfied, but the RCS hot leg temperatures are
less than the saturation temperature corresponding to the low-head (RHR) SI pump
head at minimum pump recirculation flow, the charging/SI pump can be stopped if a low-
head SI pump is running or can be started. Starting a low-head SI pump for this case
ensures that RCS subcooling will be maintained after the charging/SI pump is
stopped. (emphasis added)
The applicant incorrectly stated, however, that distractor C, Prevents a challenge to the Core
Cooling critical safety function, was also true because Q32 meant Foldout Page criteria was
met for SI Reinitiation Criteria. However, as noted earlier, Q32 asked the reason for specific
actions associated with an EOP-1.2 RNO step where, the operator starts an RHR pump and
then stops the SI pump. The FOLDOUT PAGE FOR EOP-1.2 UNIT 1 for SI REINITIATION
CRITERIA, conversely, provided instructions to manually start ECCS pumps as necessary to
restore RCS subcooling and RZR level, but it does not stop any pump.
POST-EXAMINATION COMMENT, EVALUATION, AND RESOLUTION
Nothing on the Foldout Page stops the SI pump as provided in the question stem, nor was it a
consequence of other stem conditions. NUREG-1021, Appendix E, Section 7 states, in part,
When answering a question, do not make assumptions regarding conditions that are not
specified in the question unless they occur as a consequence of other conditions that are stated
in the question. Therefore, the applicants assumption that Q32 referred to SI REINITIATION
CRITERIA on the FOLDOUT PAGE FOR EOP-1.2 UNIT 1, and its corresponding reason of
preventing a challenge to the Core Cooling critical safety function was incorrect.
The facilitys assertion that Q32s high miss rate provided evidence of a flawed question was not
supported. A high miss question could also indicate other problems, such as a training issue,
as shown by the FORM 9 - EXAM ITEM ANALYSIS for Q89, which had an even higher miss
rate, but was not challenged as a flawed question. Finally, having a similar style question,
where procedure step RNOs were used as distractors and the correct answer was based on
foldout page actions, did not a provide a valid argument that supported distractor C as correct
in Q32. Every question has different initial conditions and question stems. Applicants cannot
assume that information from a different question can be used to answer any other question on
the written examination. Each questions conditions must be considered individually (i.e., do
not make assumptions regarding conditions that are not specified in the question) per NUREG-
21, Appendix E, Section 7, as cited earlier.
Therefore, the
annotated on the answer key, was the only correct answer, and the question was considered
acceptable as administered. This challenge was denied.
SIMULATION FACILITY FIDELITY REPORT
Facility Licensee: Point Beach Nuclear Plant
Facility Docket Nos: 50-266, 50-301 and 72-005
Operating Tests Administered: July 26, 2021 through July 30, 2021
The following documents observations made by the U.S. Nuclear Regulatory Commission
examination team during the initial operator license examination. These observations do not
constitute audit or inspection findings and are not, without further verification and review,
indicative of non-compliance with Title 10 of the Code of Federal Regulations, Part 55.45(b).
These observations do not affect U.S. Nuclear Regulatory Commission certification or approval
of the simulation facility other than to provide information which may be used in future
evaluations. No licensee action is required in response to these observations.
During the conduct of the simulator portion of the operating tests, the following items were
observed:
ITEM DESCRIPTION
None
3