IR 05000280/2021301
| ML21314A567 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 11/10/2021 |
| From: | Gerald Mccoy Division of Reactor Safety II |
| To: | Stoddard D Virginia Electric & Power Co (VEPCO) |
| References | |
| 50-280/21-301, 50-281/21-301 50-280/OL-21, 50-281/OL-21 | |
| Download: ML21314A567 (24) | |
Text
November 10, 2021
Mr. Daniel Senior Vice President and Chief Nuclear Officer Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 29060
SUBJECT: SURRY NUCLEAR PLANT - NRC OPERATOR LICENSE EXAMINATI ON REPORT 05000280/2021301 and 05000281/2021301
Dear Mr. Stoddard:
During the period August 23 - 26, 2021, and October 5, 2021, th e Nuclear Regulatory Commission (NRC) administered ope rating tests to employees of y our company who had applied for licenses to operate the Surry Nuclear Plant. One R O applicant was unable to complete the operating test in August, but subsequently complet ed the operating test on October 5, 2021. The written examination was administered by y our staff on September 3, 2021. At the conclusion of the operating test, the chief exami ner discussed preliminary findings related to the operating tests with those members of your staff identified in the enclosed report.
Four Reactor Operator (RO) and three Senior Reactor Operator (S RO) applicants passed the operating test and written examination. Two SRO applicants pas sed the operating test but failed the written examination. There were fourteen post-admin istration comments concerning the written examination. These comments, and the NRC resolutio n of these comments, are summarized in Enclosure 2. A Simulator Fidelity Report is incl uded in this report as Enclosure 3.
The initial examination submittal was within the range of accep tability expected for a proposed examination. All examination changes agreed upon between the N RC and your staff were made according to NUREG-1021, Operator Licensing Examination S tandards for Power Reactors, Revision 11.
In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and its enclosures will be available electronically for public inspecti on in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRCs document system (ADAMS). ADAMS is acce ssible from the NRC Website at http://www.nrc.gov/reading-rm.adams.html (the Public Electronic Reading Room). If you have any questions concerning this letter, please contac t me at (404) 997-4551
Sincerely,
/RA/
GeraldJ.McCoy,Chief OperationsBranch1 DivisionofReactorSafety
Docket Nos: 50-280 and 50-281 License Nos: DPR-32 and DPR-37
Enclosures:
1. Report Details 2. Facility Comments and NRC Resolution 3. Simulator Fidelity Report
REGION II==
Examination Report
Docket No.: 50-280, 50-281
Report No.: 05000280/2021301 and 05000281/2021301
Enterprise Identifier: L-2021-OLL-0044
Licensee: Virginia Electric & Power Company
Facility: Surry Nuclear Plant, Units 1 and 2
Location: 5850 Hog Island Road Surry, VA 23883
Dates: Written Examination - September 3, 2021 OperatingTest-A ugust 23 - 26, 2021, and October 5, 2021
Examiners: Bruno Caballero, Chief Examiner, Senior Operations Engineer Jos eph Viera, Senior Operations Engineer Michael Kennard, Senior Operations Engineer Bernard Litkett, Reactor Engineer
Approved by: Gerald J. McCoy, Chief OperationsBranch1 DivisionofReactorSafety
Enclosure 1 SUMMARY
ER 05000280/2021301, 05000281/2021301; operating test August 23 - 26, 2021 and October 5, 2021 & written exam September 3, 2021; Surry Nuclear Plant; Operator License Examinations.
Nuclear Regulatory Commission (NRC) examiners conducted an init ial examination in accordance with the guidelines in Revision 11, of NUREG-1021, " Operator Licensing Examination Standards for Power Reactors." This examination im plemented the operator licensing requirements identified in 10 CFR §55.41, §55.43, and §55.45, as applicable.
Members of the Surry Nuclear Plant staff developed both the ope rating test and the written examination. The initial operating test, written RO examinatio n, and written SRO examination met the quality guidelines contained in NUREG-1021.
Members of the Surry Nuclear Plant training staff administered the written examination on September 3, 2021. The NRC administered the operating tests du ring the periods August 23 -
26, 2021 and October 5, 2021. Four Reactor Operator (RO) and th ree Senior Reactor Operator (SRO) applicants passed both the operating test and written exa mination. Two SRO applicants passed the operating test but failed the written examination. Seven applicants were issued licenses commensurate with the level of examination administere d.
There were fourteen post-examination comments.
No findings were identified.
REPORT DETAILS
4. OTHER ACTIVITIES
4OA5 Operator Licensing Examinations
a. Inspection Scope
The NRC reviewed the licensees examination security measures w hile preparing and administering the examinations in order to ensure compliance wi th 10 CFR §55.49, Integrity of examinations and tests.
The NRC performed an audit of license applications during the p reparatory site visit to confirm that they accurately reflected the subject applicants qualifications in accordance with NUREG-1021.
Members of the Surry Nuclear Plant training staff administered the written examination on September 3, 2021. The NRC administered the operating test during the period August 23 - 26, 2021; however, one RO applicant was unable to c omplete the operating test, but subsequently completed the operating test on October 5, 2021. The NRC examiners evaluated four Reactor Operator (RO) and five Senior Reactor Operator (SRO) applicants using the guidelines contained in NUREG-1021. Evaluations of applicants and reviews of associated documentation were perform ed to determine if the applicants, who applied for licenses to operate the Surry Nucle ar Plant, met the requirements specified in 10 CFR Part 55, Operators Licenses.
The NRC evaluated the performance or fidelity of the simulation facility during the preparation and conduct of the operating tests.
b. Findings
No findings were identified.
The NRC developed the written examination sample plan outline. Members of the Surry Nuclear Plant training staff developed both the operating test and the written examination. All examination material was developed in accorda nce with the guidelines contained in Revision 11, of NUREG-1021. The NRC determined th at the licensees written examination and operating test submittals were within t he range of acceptable quality for a proposed examinati on specified by NUREG-1021. Examination changes agreed upon between the NRC and the licensee were made per NURE G-1021 and incorporated into the final version of the examination material s.
During the on-site preparatory visit week (July 19 - 23, 2021), a near miss exam security event occurred. While the NRC exam team was using the simulato r, a member of the licensees training staff, who was not on the exam security agr eement, used the simulator door passcode and entered the simulator room. When t he training staff member entered the simulator room, he recognized that the NRC e xaminers were using the simulator and immediately left the simulator room. The lic ensee subsequently changed the simulator door passcode, placed the training staff member on the exam security agreement, and initiated a condition report (CR1177122 ). No compromise of
exam occurred because the training staff member could not see a ny scenario details and immediately left the simulator room.
Four RO applicants and three SRO applicants passed both the ope rating test and written examination. Two SRO applicants passed the operating t est but failed the written examination. Four RO applicants and three SRO applican ts were issued licenses.
Copies of all individual examination reports were sent to the f acility Training Manager for evaluation of weaknesses and determination of appropriate remed ial training.
The applicants provided five post-exam comments that contended written exam answer key changes and provided nine post-exam written exam comments that did not contend answer key changes. A copy of the final written examinations a nd answer keys, with all changes incorporated, may be accessed not earlier than October 2, 2023, in the ADAMS system (ADAMS Accession Numbers ML21302A209 and ML21302A212). A copy of the licensees post-examination comments may be accessed in the ADA MS system (ADAMS Accession Number ML21302A214.)
4OA6 Meetings, Including Exit
Exit Meeting Summary
On October 6, 2021, the NRC Chief Examiner discussed generic is sues associated with the operating test with Mr. David Wilson, Plant Manager, and me mbers of the Surry Nuclear Plant staff. The examiners asked the licensee if any o f the examination material was proprietary. No proprietary information was ident ified.
KEY POINTS OF CONTACT
Licensee personnel
David Wilson, Plant Manager Skip Irwin, Supervisor Training Joe Ford, Senior Instructor Mike Meyer, Senior Instructor Johnny Henderson, Director Safety & Licensing Don Shue, Director Engineering Adina LaFrance, Manager Licensing & Emergency Preparedness James Shell, Manager Outage & Planning Richard Philpot, Manager Organizational Effectiveness Allen Harrow, Manager Site Engineering Systems Programs Ron Herbert, Manager Engineering Design Josh LaFrance, Supervisor Nuclear Engineering Tim Catlett, Superintendent Maintenance Mike True, Engineering Technical Specialist III
NRC personnel
Brian Towne, NRC Resident Inspector
FACILITY AND APPLICANT POST-EXAMINATION COMMENTS AND NRC RESOLUTIONS
A complete text of the facility licensee and applicant post-exa mination comments can be found in ADAMS under Accession Number ML21302A214. The applicants pr ovided five post-exam comments that contended written exam answer key changes and the facility licensee concurred with three of the applicants comments. The applicants also pr ovided nine post-exam comments that did not contend written exam answer key changes, and the facility licensee did not concur with the applicants comments.
RO Question #28:
Three applicants contended that this question should be deleted from the exam because 1-AP-9.00, RCP Abnormal Conditions, contained conflicting informatio n about whether a manual reactor trip was required before stopping one RCP when the unit was not online. The facility licensee did not concur with the applicants contention.
Background
RO Question #28 was a two-part question and the answer key indi cated that Choice D was the correct answer. The first part of the question tested which pa rameter trend [thrust bearing (incorrect) or shaft vibration (correct)] would first reach its threshold value that required the RCP to be shutdown, in accordance with 1-AP-9.00. The second part of the question tested whether a manual reactor trip was required first, before stopping the R CP.
Three of the applicants contested the second part of the questi on. Five of the applicants picked the correct answer (i.e., Choice D), three applicants picked Choice C, and one applicant picked Choice A.
Enclosure 2 NRC Resolution: Applicants comment NOT accepted
The stem initial conditions stated that a unit startup was in p rogress and reactor power was being held at 1 x 10-8 amps to obtain data for reactor criticality when two abnormal RCP parameters occurred, i.e., thrust bearing and shaft vibration. The stem stated that further power ascension was suspended. Therefore, the reactor was critical, and the turbine generator was not online.
Based on the rate-of-rise during a ten-minute interval, the sha ft vibration would first require the RCP to be shutdown based on reaching its 20-mil threshold value, in accordance with annunciator procedure 1C-H5, RCP Shaft Danger, Steps 3 through 6:
1-AP-9.00 included a NOTE at the beginning of the procedure:
The applicants contended that the above 1-AP-9.00 NOTE conflict ed with Step 19.
Specifically, the applicants contended that Step 19 was guidanc e to trip the reactor (before tripping the RCP), but only if the unit was online, whereas the NOTE implied that tripping the reactor (before tripping the RCP) was required at all times the reactor was critical. Step 19, which stated:
However, the fill-in-the-blank statement referenced not only 1-AP-9.00, but it also referenced the annunciator procedures:
2) In accordance with the annunciator procedures and 1-AP-9.00, at this power level, a manual reactor trip __(2)___ required before the RCP is shutdown.
The annunciator procedure 1C-H5, RCP Shaft Danger, did include a step to trip the reactor, and the 1-AP-9.00 NOTE accommodated the annunciator procedures gui dance for the RCP shutdown.
Therefore, the only correct answer to the question was Choice D. The licensee subsequently initiated condition report 1182020 to revise AP-9.00.
RO Question #70:
Two applicants and the facility licensee contended that the ans wer key should be changed from Choice A to B because OP-23.2.1, Putting WGDT 1B on Holdup and WGDT 1A in Service, was not the correct procedure for releasing the Waste Gas Decay Tank (WGDT), i.e., the first fill-in-the-blank statement should have referred to OP-23.2.4, Release of WGDT 1B.
Background
RO Question #70 was a two-part question. The first part of the question tested whether a release of the B WGDT was permitted, and the second part of t he question tested the maximum WGDT radioactivity content allowed by Tech Spec 3.11.B, Gas Storage Tanks.
Neither the facility licensee nor any of the applicants contest ed the second part of the question.
The answer key indicated Choice A (release of the WGDT WAS permitted; 24600 curies) was the correct answer. All nine applicants missed the question and chose Answer B (i.e., release of the WGDT was NOT permitted; 24600 curies). All nine applica nts chose Choice B.
NRC Resolution: Applicants and facility licensee comment acce pted
The applicants and the facility licensee contended that the fir st fill-in-the-blank statement asked whether OP-23.2.1 permitted the release, which was not the proc edure used to release the
WGDT. The actual procedure used to release 1B WGDT was OP-23.2.4, Release of WGDT 1B. The applicants and the facility licensee contended that th e release was NOT permitted, i.e.,
Choice B was correct, because OP-23.2.1 was the wrong procedu re to use for releasing the WGDT.
The intent of the question was to test that neither the 0-WD-D9 annunciator procedure nor OP-23.2.4 precluded the release of the 1B WGDT with the conditions provided in the stem.
However, because the first fill-in-the-blank statement containe d an undetected typographical error, i.e., it incorrectly referred to OP-23.2.1 instead of OP -23.2.4, the question became different than intended.
In accordance with OP-23.2.1, release of the B WGDT _________ permitted.
[is vs is NOT]
All nine applicants chose Choice B, i.e., release is NOT perm itted, because the fill-in-the-blank statement referenced the wrong procedure. OP-23.2.1 does not i nclude steps to release the WGDT.
NUREG-1021, Section ES-403.D.1.b states, in part:
The following types of errors, if identified and adequately justified by the facility licensee or an applicant, are most likely to result in post-examination changes agreeable to the NRC:
- Unintended typographical errors in a question of on the answer key
Therefore, the applicants and facility licensees comment was accepted and the answer key was changed to Choice B as the only correct answer.
SRO Question #76:
Two applicants contended that the question should be deleted fr om the exam because there is no correct answer to the first part of the question. Specifica lly, the applicants contended that RCP Seal Leakoff Flow would not lower all the way to 0 gpm for a failure of the #3 seal. The facility licensee did not concur with the applicants contentio n.
Background
RO Question #76 was a two-part question. The first part of the question tested whether a failure of the #3 seal or a seal return line blockage had occurred base d on stem conditions. The second part of the question tested whether one running RCP woul d be sufficient to provide adequate mixing in the RCS in accordance with the Basis for Tec h 3.1.A.1, Reactor Coolant Pumps. The answer key indicated that Choice A was correct.
Four of the five applicants picked Choice A (correct answer) and one applicant picked Choice B and another applicant picked Choice D. Neither the facilit y licensee nor any of the applicants contested the second part of the question.
NRC Resolution: Applicants comment NOT accepted
After the exam review with the applicants, the facility license e ran the #3 seal failure malfunction on the plant reference simulator. RCP SEAL LEAKOFF FLOW on 1-C H-FR-1190 lowered to 0.00 gpm and RCP B SEAL PRESS on 1-CH-PI-1155A lowered to appro ximately 1150 psig, as indicated in the stem. In the case of the #3 seal failure, the reason why the 1-CH-FR-FE1190 seal leakoff flow indication lowered to 0.00 gpm was because th e leakoff flow would take the path of least resistance to the Primary Drain Transfer Pump Suc tion, instead of flowing through 1-CH-FR-FE1190 to the VCT.
Additionally, the facility licensee ran the seal return line bl ockage malfunction on the plant reference simulator. In the case of the seal return line block age, RCP SEAL LEAKOFF FLOW on 1-CH-FR-1190 also lowered to 0.0 gpm because the seal return line containing 1-CH-FR-1190 was blocked. However, in this case, the RCP SEAL PRESS on rose to full reactor pressure (i.e., 2235 psig), which was different than the inform ation provided in the stem.
Therefore, the only correct answer was Choice A.
RO Question #80:
Two applicants and the facility licensee contended that the que stion should be deleted from the exam because 0-AP-12.00, Service Water System Abnormal Conditio ns, did not include a requirement to shift charging pumps at 180F (i.e., at time 144 0). The two applicants and the facility licensee contended that the AP-12.00 Step 3 Response N ot Obtained (RNO) Column did not require a charging pump shift; they contended that the word s as necessary in the RNO meant that shifting pumps was not mandatory.
Background
RO Question #70 was a two-part question. The first part of the question tested when the operating charging pump was required to be shifted in accordanc e with 0-AP-12.00. The second part of the question tested the latest time the unit was required to be in cold shutdown.
Neither the applicants nor the facility licensee contested the second part of the question. The answer key indicated Choice D (180F at time 1440; Thursday a t 2100) was the correct answer. All five applicants missed the question; four applicant s picked Choice B and one applicant picked Choice C.
NRC Resolution: Applicants comment NOT accepted
The first part of the question was a fill-in-the-blank statemen t:
In accordance with 0-AP-12.00, the earliest time the operating charging pump is required to be shifted is ___(1)___.
[1440 vs 1450]
(Time 1440 was the time bearing temperatures were 180 F; time 1450 they were 185F.)
0-AP-12.00 included the following Caution, Note, and Step 3 RNO :
The CAUTION said that a charging pump should be stopped if its bearing temperature reached 185F. However, Step 3 required shifting the charging pump at 180F. Shift meant to first start a pump that was not previously running, and then stop the charging pump whose temperature had risen to 180F. The phrase as necessary mean t that shifting pumps may be required multiple times using different pumps, i.e., as many ti mes as was necessary to keep one of the charging pumps running with its bearing temperature less than 180F.
Therefore, because 0-AP-12.00 Step 3 was a procedure requiremen t, the only correct answer to the question was Choice D.
SRO Question #99:
Three applicants and the facility licensee contended that the a nswer key should be changed from Choice D to C because the NOTE in 1-E-0, Reactor Trip or Safety Injection (SI),
Attachment 1, System Alignment Verification, was no longer appl icable in 1-E-1, Loss of Reactor or Secondary Coolant. Specifically, the applicants and facility licensee contended that subsequent SI signals were not allowed to be reset using Attach ment 1, Step 12, after 1-E-1 had been entered.
Background
SRO Question #99 was a two-part question. The first part of th e question tested whether the crew was permitted to reset SI using the guidance of 1-E-0, Att achment 1, System Alignment Verification, while in 1-E-1. The second part of the question tested the required procedural path to transition to 1-ECA-1.2, LOCA Outside Containment, after the SRO identified that an incorrect transition to 1-E-1, had occurred. That is, 1-ES-0.0, Rediagnosis, was required.
Neither the applicants nor the facility licensee contested the second part of the question. The answer key indicated that Choice D was the correct answer.
Three applicants picked Choice C and two applicants picked Ch oice D.
NRC Resolution: Applicants and facility licensee comment acce pted
1-E-0, Step 5 required the implementation of Attachment 1 when SI was actuated. Attachment 1 was an alignment verification following an SI actuation. The second bullet in the NOTE preceding Step 12 was:
Neither the NOTE nor Step 12 were continuous action steps, i.e., there was no asterisk (*)
designator. OP-AP-104, Emergency and Abnormal Operating Proced ures, Section 3.4.1 stated pertaining to NOTES:
The applicants and facility licensee contended that, unlike con tinuous action steps, NOTES do not carry forward into subsequent procedures.
OP-AP-104, Section 3.4.6 included guidance for continuous actio n steps in the following note:
However, the 1-E-0 Attachment 1 NOTE was not a continuous actio n step; therefore, it was not applicable when 1-E-1 was entered. Therefore, the applicants and facility licensees comment was accepted and the answer key was changed to Choice C as th e only correct answer. The licensee subsequently initiated condition report 1182020 to rev ise E-0, Attachment 1.
The applicants also provided the following nine comments; howev er, none of these applicant comments contended answer key changes, and the facility license e did not concur with these applicant comments.
RO Question #5:
One applicant commented that 1-AP-9.00, RCP Abnormal Conditions, required a reactor trip with a loss of seal injection.
NRC Resolution: No answer key change was required. 1-AP-8.00, Loss of Normal Charging Flow, was referenced in the second fill-in-the-blank (not AP-9. 00), and both 1-AP-8.00 and 1-AP-9.00 required a manual reactor trip only when all seal injec tion and thermal barrier cc flow was lost. The stem of the question did not include any informa tion related to a loss of thermal barrier cc flow.
RO Question #14:
Two applicants commented that the operating point on the Genera tor Capability Curve (1-DRP-003, Attachment 52) looked to be right on the curve.
NRC Resolution: No answer key change was required. The point where 900 MWe and +350 VARS intersected was a reasonable distance outside the 60-psig generator hydrogen curve, the curve provided to the applicants as a reference is the actual c urve used by operators and was enlarged for the applicants.
RO Question #21:
One applicant commented that the underlined portion of the firs t fill-in-the-blank statement (i.e.,
without I&C support) led them to the wrong answer.
NRC Resolution: No answer key change was required. 1-AP-4.00, Nuclear Instrumentation Malfunction, contained steps for operators to shift the audible containment count rate without I&C support, whereas other 1-AP-4.00 steps required I&C support.
RO Question #23:
One applicant commented by asking a question how do you know t he affected area, and what does health physics do when sent out to the affected area?
NRC Resolution: No answer key change was required. Step 10 of 0-AP-5.20, Radiation Monitor System Ventilation Vent High Alarm, provided guidance f or the stem situation where the source of ventilation vent activity was unknown. Step 10 requir ed placing one area at a time on the filtered exhaust system to identify a change in ventilation activity.
RO Question #24:
One applicant commented by asking a question why are we not us ing 0-AP-22.00; the first fill-in-the-blank statement referenced annunciator procedures.
NRC Resolution: No answer key change was required. Annunciato r procedure 0-RM-D3, 1-RM-FI-153 HIGH, and 0-AP-22.0 both required fuel building evacu ation.
RO Question #48:
One applicant commented by stating I confused the output break er with excitation status.
NRC Resolution: No answer key change was required. The second fill-in-the-blank statement tested the applicants knowledge on how excitation was affected by a loss of the Unit 2 A DC Bus. Specifically, second fill-in-the-blank statement was If an automatic turbine trip occurred, Main Generator excitation ________ remain energized with no operator action.
RO Question #67:
One applicant commented that the second part of the question te sted a procedure change that the class did not have and was minutia.
NRC Resolution: No answer key change was required. The second part of the question tested the October 2019 version of SUADM-O-26, Administrative Control of Operational Components, which was in effect before the class started. The second fill-in-the-blank statement was It
_______ acceptable for an operator assigned to the Fire Team to also have an administrative control function. Lesson Plan RO/SRO/SROUTP-SDS-2, Administrative Procedures, identified SUADM-O-26 as a Tier 1 procedure, which required an in-depth knowledge of the procedure, and included Learning Objective A, which was 1) purpose of the procedure, 2) operations personnel responsibilities, 3) who is required to authorize spe cific plant operations or process changes, 4) knowledge items applicable to operators.
SRO Question #79:
One applicant commented that during the exam they requested the emergency action levels (EALs) basis document because there was not enough information to rule out a General Emergency (GE).
NRC Resolution: No answer key change was required. For a GE c lassification to be required, the determination that long-term RCS heat removal capability w as not likely to be established and maintained per procedure was needed. The stem indicated th at the event had been in
progress for only seventeen minutes, and the EAL basis document credited flex equipment and flex strategy towards maintenance of long-term RCS heat removal capability.
SRO Question #94:
One applicant commented that LI-AA-700, Fatigue Management and Work Hour Limits for Covered Workers, contained guidance that the Shift Manager had authority to approve a waiver.
NRC Resolution: No answer key change was required. The NOTE pr eceding Step 3.12.5 defined site senior-level manager as the site vice president, plant manager, or Director Nuclear Station Safety and Licensing.
SIMULATOR FIDELITY REPORT
Facility Licensee: Surry Nuclear Plant
Facility Docket No.: 05000280, 05000281
Operating Test Administered: August 23 - 26, 2021 and October 5, 2021
This form is to be used only to report observations. These obs ervations do not constitute audit or inspection findings and, without further verification and re view in accordance with Inspection Procedure 71111.11 are not indicative of noncompliance with 10 CFR 55.46. No licensee action is required in response to these observations.
During the validation of a portion of the operating test, the e xaminer observed the following:
Item Description
1. Feedwater Regulating Valve Controllers OK status light was not illuminated. Simulator Work Order # 202110041000
Enclosure 3