IR 05000483/2023301
| ML23313A115 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 11/09/2023 |
| From: | Heather Gepford NRC/RGN-IV/DORS/OB |
| To: | Diya F Ameren Missouri |
| References | |
| IR 2023301 | |
| Download: ML23313A115 (1) | |
Text
November 09, 2023
SUBJECT:
CALLAWAY PLANT, UNIT 1 - NRC EXAMINATION REPORT 05000483/2023301
Dear Mr. Diya:
On November 1, 2023, the U.S. Nuclear Regulatory Commission (NRC) completed an initial operator license examination at Callaway Plant, Unit 1. The enclosed report documents the examination results and licensing decisions. The preliminary examination results were discussed on September 14, 2023, with you and other members of your staff. A telephonic exit meeting was conducted on November 1, 2023, with Mr. K. Scott, Site Vice President, who was provided the NRC licensing decisions.
The examination included the evaluation of three applicants for reactor operator licenses, five applicants for instant senior reactor operator licenses, and one applicant for an upgrade senior reactor operator license. The license examiners determined that eight of the nine applicants satisfied the requirements of 10 CFR Part 55, and the appropriate licenses have been issued.
There were four 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. 05000483 License No. NPF-30 Enclosures:
1. Examination Report 05000483/2023301 2. NRC Post-Examination Comment Resolution Electronic distribution via LISTSERV Signed by Gepford, Heather on 11/09/23
ML23313A115 SUNSI Review: ADAMS: Non-Publicly Available Non-Sensitive Keyword:
By: tjf Yes No Publicly Available Sensitive NRR-079 OFFICE RIV/DORS/OB RIV/DORS/OB RIV/DORS/OB RIV/DORS/OB C:OB NAME TFarina KMurphy RWilliams DYou HGepford SIGNATURE /RA/ /RA/ /RA/ /RA/ /RA/
DATE 11/07/23 11/08/23 11/08/23 11/08/23 11/09/23
U.S. NUCLEAR REGULATORY COMMISSION Examination Report
Docket Number: 05000483
License Number: NPF-30
Report Number: 05000483/2023301
Enterprise Identifier: L-2023-OLL-0029
Licensee: Union Electric Company
Facility: Callaway Plant
Location: Steedman, Missouri
Examination Dates: September 11, 2023, to November 1, 2023
Examiners: T. Farina, Senior Operations Engineer (Chief Examiner)
K. Murphy, Operations Engineer R. Williams, Operations Engineer D. You, Operations Engineer
Approved By: Heather J. Gepford, Ph.D., Chief Operations Branch Division of Operating Reactor Safety
Enclosure 1 SUMMARY
Examination Report 05000483/2023301; September 11 to November 1, 2023; Callaway Plant, Unit 1; Initial Operator Licensing Examination Report
The NRC examiners evaluated the competency of three applicants for reactor operator licenses, five applicants for instant senior reactor operator licenses, and one applicant for an upgrade senior reactor operator license at Callaway Plant, Unit 1.
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 September 20, 2023. The NRC examiners administered the operating tests from September 11-14, 2023.
The NRC examiners determined that eight of the nine applicants satisfied the requirements of 10 CFR Part 55, and the appropriate licenses have been issued.
A. NRC-Identified and Self-Revealing Findings
No findings were identified, but the Senior Reactor Operator portion of the written examination was determined to be outside of the acceptable quality range as identified in NUREG-1021.
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 two 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.
The Senior Reactor Operator (SRO) portion of the written examination was determined to be outside of the acceptable quality range as identified in NUREG-1021. For the SRO portion of the exam, seven questions (28 percent) required significant modification, replacement, or deletion as a result of the NRC examiner review and post-exam comment resolution.
Flaws included not testing of knowledge only required of the SRO, insufficient question-to-randomly selected KA statement match, questions containing more than one correct answer, and questions with no correct answer. Two SRO questions (and one RO question) required key changes or deletion as a result of licensee-and applicant-submitted post-examination comments. Future written examination submittals need to incorporate lessons learned.
.3 Operator Knowledge and Performance
a. Scope
On September 20, 2023, the licensee proctored the administration of the written examinations to all nine applicants. The licensee staff graded the written examinations, analyzed the results, and presented their analysis and post-examination comments to the NRC on October 2, 2023.
The NRC examination team administered the various portions of the operating tests to all applicants from September 11-14, 2023.
b. Findings
No findings were identified.
Eight of the nine applicants passed the written examination and all parts of 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 dynamic scenario section of the operating tests. The applicants displayed a weakness addressing technical specifications for a loss of B charging pump. Post-examination analysis revealed ten generic weaknesses associated with applicant performance on the written examination. Specifically, the following ten questions had a miss-rate of 50 percent or higher: Questions 8, 14, 18, 25, 58, 76, 78, 83, 88, and 89. These weaknesses were captured in the licensees corrective action program as Condition Reports 2023-06586, -06587, -06588, -07307, -07308, 07310,
-07311, -07312, -07313, and -07314. 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.
b. Findings
No findings were identified.
EXIT MEETINGS AND DEBRIEFS
Exit Meeting Summary
The chief examiner presented the preliminary examination results to Mr. F. Diya, Senior Vice President, and other members of the staff on November 14, 2023. A telephonic exit was conducted on November 1, 2023, between Mr. T. Farina, chief examiner, and Mr. K. Scott, Site Vice President.
The licensee did not identify any information or materials used during the examination as proprietary.
ADAMS DOCUMENTS REFERENCED
Accession No. ML23311A296 - FINAL WRITTEN EXAMS Accession No. ML23311A298 - FINAL OPERATING TEST Accession No. ML23305A094 - POST-EXAMINATION ANALYSIS-COMMENTS
NRC Resolution to Callaway Plant Post-Examination Comments
The complete text of the licensee's post-examination analysis and comments can be found in ADAMS under Accession Number ML23305A094.
RO QUESTION # 56
Enclosure 2 The facility and an applicant argue that the keyed answer B: LICENSEE COMMENT:
Annunciator 32D, PZR Hi Lev Dev Htrs On, alarms and PZR backup heaters turn on, does not correctly describe the response of the pressurizer to a rapid down power from 100 to 80 percent, at 3 percent per minute. Rather, they argue that the correct answer should be D:
Annunciator 33C, PZR Press Lo Htrs On, alarms and PZR backup heaters turn on. The basis for the argument is that pressurizer program level varies with RCS average coolant temperature, and control rods will automatically insert when RCS average coolant temperature exceeds reference RCS temperature by 1.5 degrees F. The combination of these two effects will limit the rise in RCS average coolant temperature and thereby limit the pressurizer insurge during the down power, preventing pressurizer level from exceeding 5 percent deviation from program level, which is the setpoint for alarm 32D, PZR Hi Lev Dev Htrs On.
The facility conducted scenario-based testing of this evolution on the plant-referenced simulator.
The results of the testing demonstrated that control rods were able to limit the rise in RCS average coolant temperature, and annunciator 32D, PZR Hi Lev Dev Htrs On, never alarmed during 16 minutes of the transient. However, while the plant was stabilizing itself with rods stepping in to return RCS average coolant temperature to its new lower reference temperature, the pressurizer outsurge caused Annunciator 33C, PZR Press Lo Htrs On, to alarm at around 2203 psig, which caused pressurizer backup heaters to energize for a different reason than the question was originally keyed for. Answer D was intended to be a distractor, and reads:
Annunciator 33C, PZR Press Lo Htrs On, alarms and PZR backup heaters turn on. Based on
the observed plant response, this would be the actual correct answer, thereby the key should be changed to accept D as correct, and B as incorrect.
NRC RESOLUTION: The NRC agrees with the licensees recommendation. Based on a review of the data contained in the scenario-based testing package, conducted in accordance with station procedure TDP-IS-00002 Appendix C, Simulator Scenario Based Testing and Documentation, the NRC agrees that D, Annunciator 33C, PZR Press Lo Htrs On, alarms and PZR backup heaters turn on is the correct answer, and that B, Annunciator 32D, PZR Hi Lev Dev Htrs On, alarms and PZR backup heaters turn on is incorrect. Pressurizer level does not rise enough beyond program level during this rapid down power to cause Annunciator 32D to alarm, but pressurizer pressure does lower enough to cause Annunciator 33C to alarm and energize pressurizer backup heaters. The key was edited to reflect the change.
SRO QUESTION # 81
The licensee stated that due to an applicant question during exam LICENSEE COMMENT:
administration regarding the flow rate to each SG (Is the 20,000 lbm/hr [AFW flow] due to operator action or due to equipment malfunctions? Where in FR-H.1 are we?), it was determined that further analysis of this question was required.
Part (1) of the question asked, Per FR-H.1 and based on the above conditions, what action(s)
should the crew take NEXT? The licensee stated that the intent of part (1) was for the applicant to recognize that with SG levels at 25%, RCS Bleed and Feed was required per continuous action Step #2.c, which sends the applicant to steps 12 and 13, initiating safety injection and opening both PORVs. This expectation was keyed as the correct answer, specifically, Initiate safety injection and open both PZR PORVs. However, the licensee subsequently commented that before these steps would be performed, the crew would have to complete Step #2.b, Stop all RCPs. Stopping RCPs was not an available answer for the applicants to select; therefore, the licensee contends that there was no correct answer available for part (1).
The licensee further challenged the psychometric attributes of part (1), specifically that the difference between the correct answer and the distractor was of little discriminatory value. The part (1) distractor states, Feed all SG at maximum rate until core exit TCs lower. The licensee stated that this distractor was based on Foldout Page item #3, SG Feed Flow Restrictions FOLLOWING RCS Bleed and Feed Criteria, and that the actions of the distractor would always be performed after performing the actions of the correct answer. The licensee contends that simply knowing the procedural order of what comes first, establishing RCS Bleed and Feed or feeding all SG at maximum rate, was of little discriminatory value.
Due to part (1) having no listed correct answer and being psychometrically flawed, the licensee recommended deleting Question 81 from the exam.
NRC RESOLUTION: The NRC agrees that there is no correct answer to part (1) of this question. The question as-written started the applicants at step 1 of FR-H.1 (The crew has just entered FR-H.1). The facility proctor, in response to an applicants question, provided additional information to all applicants that the crew was actually at Step 3 of FR-H.1. This changed the stem of the question by placing the applicant at a different place in the procedure but made no material difference to the fact that there was no correct answer. The facility proctor explained to the chief examiner that the intent of this additional information was to strengthen the credibility of part (2)s correct answer, since EOP Addendum 38 is only mentioned by name in Step 3; if the applicants started at Step 1, then they would reach guidance to Continue attempts to establish secondary heat sink in at least one SG in step 22, which does include Non Safety aux feedwater flow as one option, but does not specifically reference EOP Addendum 38 by name. The proctor stated that the expectation with the new information was that the applicants should recognize that even though they were currently at Step 3, with steam generator wide range levels having lowered to 25 percent the procedure reader should go back to Step 2, which was a continuous action step: Check if RCS Bleed and Feed - REQUIRED: a. SG WIDE RANGE level in any three SGs - LESS THAN 27%.With this criterion being met, Step 2.b directs STOP all RCPs, and step 2.c directs Go to Step 12. Steps 12, 13, and 14 then direct actuating safety injection and opening pressurizer PORVs to establish RCS Bleed and Feed, which was the keyed correct answer for part (1). However, since the stem asked, What action should the crew take NEXT, the correct answer under these conditions would be stopping all RCPs per Step 2.b, which was not provided in the answer choices. Therefore, there is no correct answer provided for part (1).
The licensee separately argues that part (1) is psychometrically flawed because knowledge of procedure step order offers little or no discriminatory value, since maximizing feed rate will always be performed after meeting RCS Bleed and Feed criteria. NUREG 1021, Rev. 12, ES-4.4 states, The NRC will not accept examination changes for the following types of question errors identified after examination administration: a question that tests minutiae, even though the facility licensee and the NRC previously agreed that the question did not test minutiae.
Since the facility proposed that this was an acceptably discriminatory question when submitted, a low discriminatory value (i.e., minutiae) identified post-exam would not be a justifiable reason for deleting the question.
The NRCs disagreement regarding the discriminatory value of part (1) notwithstanding, there was no correct answer to the question provided to the applicants, and Question 81 was deleted from the exam.
SRO QUESTION # 83
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An applicant challenged that the keyed answer for part (2) of the LICENSEE COMMENT:
question is incorrect. Part (2) asks What action is required BEFORE the Shift Manager can authorize installation of the TM? with the keyed answer of SRO review complete and documented on CA4691, Comment/Impact Review Only. The distractor for part (2) adds an additional requirement that Onsite Review Committee (ORC) approval is required prior to TM installation. While APA-ZZ-00091, On-Site review Committee, step 4.2.3.a.18 states that ORC Safety Related Design changes shall be reviewed by the ORC and approved by the Senior Director, Nuclear Operations, it does not specify if this approval is before or after TM installation; this information is found in APA-ZZ-00605. Specifically, APA-ZZ-00605 step 4.4.2 states, A TM requiring ORC approval may be installed prior to ORC approval provided the TM is ORC approved by the close of the next working day, as discussed in the explanation of the questions correct answer. However, APA-ZZ-00605 Step 4.4.1 does direct the implementation of IP-ENG-001, Standard Design Process. IP-ENG-001 defines temporary modification as A short-term alteration made to systems, structures, or components that is not controlled by procedure or work order instructions, and is evaluated via a temporary Commercial Change, Design Equivalent Change or Design Change. Therefore, the Design Equivalent Change process flow, Attachment 5, will apply. Per the flowchart on Attachment 5, page 1, Steps 3.3.13 and 3.3.14, review by Plant Review Board then Plant Manager or Site VP approval occur before the approved package is ready for WO planning and implementation. It therefore appears that IP-ENG-001 requires ORC approval before implementation. Note per APA-ZZ-00091, the plant manager is the ORC chairperson which implies the plant review board is the ORC.
Due to the conflicting procedural guidance of APA-ZZ-00605 vice APA-ZZ-00091 with IP-ENG-001, Callaway recommends that this question be removed from the exam.
CR 202306570 was generated to revise APA-ZZ-00605 to remove the conflict.
NRC RESOLUTION: The NRC disagrees with the facilitys recommendation. Implementation of Temporary Modifications (TMs) at Callaway is covered by several procedures, some station-specific and some industry standards. Of note, APA-ZZ-00605, Temporary System Modifications, revision 45 is directly applicable. Section 4.4.1.b of this procedure, Engineering Approval Package Generation, directs the responsible engineer to PROCESS Temporary Modification Package (MP) in accordance with EDP-ZZ-04600, Engineering Change Control, and IP-ENG-001, Standard Design Process. EDP-ZZ-04600 Attachment 5, Site Nomenclature, explicitly states that the IP-ENG-001 term Plant Review Board is equivalent to the Callaway term Onsite Review Committee (ORC). Procedure APA-ZZ-00091, On-Site Review Committee, step 4.2.3.a.18 states that ORC Safety Related Design changes shall be reviewed by the ORC and approved by the Senior Director, Nuclear Operations. As mentioned by the licensee however, this guidance does not explicitly specify if ORC approval must be obtained before or after TM installation. APA-ZZ-00605 section 4.4.2.a, Engineering Approval Package Submittal, requires the responsible engineer to ENSURE ORC has approved the TM package in accordance with APA-ZZ-00091, Onsite Review Committee, if necessary.
Amplifying guidance is then immediately provided: A TM requiring ORC approval may be installed prior to ORC approval provided the TM is ORC approved by the close of the next working day.
Industry Standard IP-ENG-001, Standard Design Process (EB-17-06), revision 1, was developed by the industry in support of the Delivering the Nuclear Promise initiative. Its stated purpose is, in part, to provide the Standard Design Process for the Nuclear Industry.
The purpose further states that [t]his procedure is to be utilized in conjunction with Utility-specific procedures that implement interfacing processes that support the design control, engineering change and configuration management processes. Section 4.1, Precautions and Limitations, states that, The Utility-specific interface procedures should clearly delineate those transition areas and provide clear guidance to users where necessary. Those Utility-specific procedures should not redefine, reinterpret, or modify the intent of this procedure.
The licensee stated that Attachment 5 of IP-ENG-001, Design Equivalent Change Package, is the appropriate process to follow for processing the temporary modification described in Question 83. Step 3.3.13 of this flowchart states that the change package receives, Review by Plant Review Board (if required), followed by Step 3.3.14, Plant Manager or Site VP approval (if required), and finally step 3A, Approved/Issued Design Equivalent Change Package (ready for WO planning and implementation). The modifier, if required, in steps 3.3.13 and 3.3.14 is important to the evaluation of this question. The amplifying guidance of step 3.3.13 in Attachment 5 states, If the Utility-specific procedures require the Design Equivalent Change to be reviewed by the Plant Review Board (PRB) or equivalent, the Responsible Engineer (RE)
shall schedule the review Likewise, step 3.3.14 guidance states, If the Utility-specific procedures require the Design Equivalent Change to be reviewed and approved by the Site Vice President or Plant Manager, the Responsible Engineer shall obtain the approval signature
Industry Standard IP-ENG-001 states throughout that it is to be used in conjunction with Utility-specific procedures. The order of workflow in steps 3.3.13 and 3.3.14 state that ORC review and Plant Manager approval take place before Temporary Modification approval only If the Utility-specific procedures require the review. Utility-specific procedure APA-ZZ-00605, which directs the use of EDP-ZZ-04600 and IP-ENG-001 in step 4.4.1.b, provides an explicit allowance in step 4.4.2.a that A TM requiring ORC approval may be installed prior to ORC approval provided the TM is ORC approved by the close of the next working day.
Therefore in summary, industry standard IP-ENG-001 defers to the Utility-specific procedure to determine if ORC review and approval is required before Temporary Modification installation; Callaway-specific procedures allow for Temporary Modification installation up to one working day before ORC approval. Based on that, it must be considered that per Callaways approved procedures, ORC approval is not REQUIRED before the Shift Manager can authorize the installation of the Temporary Modification. Therefore, the NRC has determined that the correct answer to Question 83 remained C.
The purpose of this assessment was to evaluate whether the question should be deleted due to conflicting guidance and/or the answer rekeyed. As such, the assessments scope was limited to evaluating whether Callaways station procedures, as currently written, required obtaining ORC approval prior to the shift manager authorizing installation of a temporary modification.
SRO QUESTION # 89
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15 The licensee recommended changing the correct answer to C (78 LICENSEE COMMENT:
hours) vice B (7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />). The licensee asserted that LCO 3.6.6 Condition A (one containment spray train INOP) and Condition C (one containment cooling train INOP) are entered separately and tracked separately per the guidance of Tech Spec example 1.3-3, and because they are separate functions, none of the entry criteria for LCO 3.0.3 apply. Specifically:
- When an LCO is not met and the associated ACTIONS are not met
- An associated ACTION is not provided, or
- If directed by the associated ACTIONS
The licensee asserted that although there is no single action statement covering the condition of one train of each function inoperable, the individual action statements for Conditions A and C are adequate when tracked concurrently, therefore an associated combination of actions IS provided, and LCO 3.0.3 does not apply. Therefore, containment spray is the limiting function, and the longest permissible time allowed before the plant must be in Mode 3 is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (condition A) + 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (condition D) = 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br />, which is answer C.
NRC RESOLUTION: The NRC concurs with the licensees recommendation. Technical Specification 3.0.3 basis states:
LCO 3.0.3 establishes the actions that must be implemented when an LCO is not met and:
a. An associated Required Action and Completion Time is not met and no other Condition applies; or b. The condition of the unit is not specifically addressed by the associated ACTIONS. This means that no combination of Conditions stated in the ACTIONS can be made that exactly corresponds to the actual condition of the unit.
Under the conditions of Question 89, Tech 3.6.6 does provide a combination of actions that can be made to exactly correspond to the actual condition of the unit. Specifically, Condition A can be entered for one containment spray train inoperable, and Condition C can be entered for one containment cooling train inoperable. If one train of containment spray cannot be restored in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, then Condition B applies, and the unit must be in Mode 3 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> later. This aligns with Technical Specification example 1.3-3, as the licensee noted.
Given the above, the correct answer for Question 89 was changed to C.
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