IR 05000327/2022301
| ML22210A126 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 07/21/2022 |
| From: | Tom Stephen Division of Reactor Safety II |
| To: | Jim Barstow Tennessee Valley Authority |
| References | |
| 50-327/22-01, 50-328/22-01 50-327/OL-22, 50-328/OL-22 | |
| Download: ML22210A126 (20) | |
Text
July 21, 2022
SUBJECT:
SEQUOYAH NUCLEAR PLANT - NRC OPERATOR LICENSE EXAMINATION REPORT 05000327/2022301 AND 05000328/2022301
Dear Mr. Barstow:
During the period May 2 - 11, 2022, the Nuclear Regulatory Commission (NRC) administered operating tests to employees of your company who had applied for licenses to operate the Sequoyah Nuclear Plant. At the conclusion of the tests, the examiners discussed preliminary findings related to the operating tests with those members of your staff identified in the enclosed report. The written examination was administered by your staff on May 18, 2022.
Five Reactor Operator (RO) and eleven Senior Reactor Operator (SRO) applicants passed both the operating test and written examination. One RO and one SRO applicant failed the written examination. There were two post-administration comments concerning the operating test, and four post-administration comments concerning the written examination. These comments, and the NRC resolution of these comments, are summarized in Enclosure 2. A Simulator Fidelity Report is included in this report as Enclosure 3.
The initial examination submittal was within the range of acceptability expected for a proposed examination. All examination changes agreed upon between the NRC and your staff were made according to NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 12.
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 inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRCs document system (ADAMS). ADAMS is accessible 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 contact me at (404) 997-4703.
Sincerely,
/RA/
Thomas A. Stephen, Chief
Operations Branch 1
Division of Reactor Safety
Docket Nos: 50-327, 50-328 License Nos: DPR-77, DPR-79
Enclosures:
1. Report Details 2. Facility Comments and NRC Resolution 3. Simulator Fidelity Report
REGION II==
Examination Report
Docket No.:
05000327, 05000328
License No.:
Report No.:
05000327/2022301 and 05000328/2022301
Enterprise Identifier: L-2022-OLL-0030
Licensee:
Tennessee Valley Authority
Facility:
Sequoyah Nuclear Plant
Location:
Soddy-Daisy, TN
Dates:
Operating Test - May 2 - 11, 2022
Written Examination - May 18, 2022
Examiners:
M. Meeks, Chief Examiner, Senior Operations Engineer
C. Zoia, Senior Operations Engineer J. Bundy, Operations Engineer A. Goldau, Operations Engineer K. Kirchbaum, Operations Engineer J. Correll, Operations Engineer (observation/training only)
Approved by:
Thomas A. Stephen, Chief
Operations Branch 1
Division of Reactor Safety
SUMMARY
ER 05000327/2022301, 05000328/2022301; May 2 - 11, 2022 & May 18, 2022; Sequoyah
Nuclear Plant; Operator License Examinations.
Nuclear Regulatory Commission (NRC) examiners conducted an initial examination in accordance with the guidelines in Revision 12 of NUREG-1021, "Operator Licensing Examination Standards for Power Reactors." This examination implemented the operator licensing requirements identified in 10 CFR §55.41, §55.43, and §55.45, as applicable.
Members of the Sequoyah Nuclear Plant staff developed both the operating tests and the written examination. The initial operating test, written Reactor Operator (RO) examination, and written Senior Reactor Operator (SRO) examination submittals met the quality guidelines contained in NUREG-1021.
The NRC administered the operating tests during the period May 2 - 11, 2022. Members of the Sequoyah Nuclear Plant training staff administered the written examination on May 18, 2022.
Five RO and eleven SRO applicants passed both the operating test and written examination.
Fifteen (15) applicants were issued licenses commensurate with the level of examination administered.
There were a total of six post-examination comments.
No findings were identified.
REPORT DETAILS
OTHER ACTIVITIES
4OA5 Operator Licensing Examinations
a. Inspection Scope
The NRC evaluated the submitted operating test by combining the scenario events and JPMs in order to determine the percentage of submitted test items that required replacement or significant modification. The NRC also evaluated the submitted written examination questions (RO and SRO questions considered separately) in order to determine the percentage of submitted questions that required replacement or significant modification, or that clearly did not conform with the intent of the approved knowledge and ability (K/A) statement. Any questions that were deleted during the grading process, or for which the answer key had to be changed, were also included in the count of unacceptable questions. The percentage of submitted test items that were unacceptable was compared to the acceptance criteria of NUREG-1021, Operator Licensing Standards for Power Reactors.
The NRC reviewed the licensees examination security measures while preparing and administering the examinations in order to ensure compliance with 10 CFR §55.49, Integrity of examinations and tests.
The NRC performed an audit of license applications during the preparatory site visit in order to confirm that they accurately reflected the subject applicants qualifications in accordance with NUREG-1021.
The NRC administered the operating tests during the period May 2 - 11, 2022. The NRC examiners evaluated six Reactor Operator (RO) and twelve Senior Reactor Operator (SRO) applicants using the guidelines contained in NUREG-1021. Members of the Sequoyah Nuclear Plant training staff administered the written examination on May 18, 2022. Evaluations of applicants and reviews of associated documentation were performed to determine if the applicants, who applied for licenses to operate the Sequoyah Nuclear 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 Sequoyah Nuclear Plant training staff developed both the operating tests and the written examination. All examination material was developed in accordance with the guidelines contained in Revision 12 of NUREG-1021. The NRC examination team reviewed the proposed examination. Examination changes agreed upon between the NRC and the licensee were made per NUREG-1021 and incorporated into the final version of the examination materials.
The NRC determined, using NUREG-1021, that the licensees initial examination submittal was within the range of acceptability expected for a proposed examination.
Several errors were identified by the NRC examiners during the administration of the operating test. After administering the A3 Radiological Controls topic administrative JPM to the RO applicants, the examiners determined that the approved final JPM (which had replaced an unsatisfactory draft A3 JPM) was invalid. The examiners applied the guidance in NUREG-1021 ES-3.5 B(2), and received Branch Chief approval to administer a different Radiological Controls administrative JPM to all RO applicants.
See the post-exam comments section for further details. The examiners identified two simulator discrepancies to the facility which are documented below in the simulator fidelity section of this report. Finally, during the onsite validation week, the NRC examiners requested modifications be made to the first event of Scenario 5, a normal evolution involving main feed pump startup, the facility licensee did not ensure the simulator load for this evolution was correctly modified. This resulted in the first team to run scenario 5 being unable to complete the first evolution. The facility licensee was able to correct the issue with the simulator load, and the other two teams to run the scenario were able to successfully complete the simulated event as intended. Due to the issues with the first run of Scenario 5, the facility licensee was not able to save simulator parameters associated with that teams performance.
Five RO applicants and eleven SRO applicants passed both the operating test and written examination. One RO applicant and one SRO applicant passed the operating test but did not pass the written examination. Four RO applicants and eleven SRO applicants were issued licenses. Issuance of the license for one RO applicant has been delayed pending receipt of additional information. Details concerning the need for additional information has been sent to the individual applicant(s) and the facility licensee.
During an informal debrief and the exit meeting, the NRC examiners discussed several issues identified as generic weaknesses discovered during the operating test administration. The examiners provided the facility licensee training staff multiple examples of applicant operators with verbal communications issues, specifically with reading and repeat-backs of procedural NOTEs and CAUTION statements. The examiners also discussed inconsistent Alarm Response procedural usage by the control board operators, as well as inconsistent Auxiliary Feedwater (AFW) control, especially understanding integrated plant response following automatic actuations of the AFW system during accident conditions.
Copies of all individual examination reports were sent to the facility Training Manager for evaluation of weaknesses and determination of appropriate remedial training.
The licensee submitted two post-examination comments concerning the operating test and four comments concerning the written examination. A copy of the final written examination and answer key, with all changes incorporated, [and the licensees post-examination comments] may be accessed not earlier than [Month XX, XXXX--two years after administration of the written exam, date must be Monday-Friday], in the ADAMS system (ADAMS Accession Number(s) MLXXXXXXXXX [and MLXXXXXXXXX].
4OA6 Meetings, Including Exit
Exit Meeting Summary
On May 13, 2022, the NRC examination team discussed generic issues associated with the operating test with Mr. C. Reneau, Sequoyah Plant Manager, and other members of the Sequoyah Nuclear Plant staff. The examiners asked the licensee if any of the examination material was proprietary. No proprietary information was identified.
KEY POINTS OF CONTACT
Licensee personnel
B. Buch, Sequoyah Operations Training A. Forsha, Sequoyah Training Examination Development G. Garner, Director of Site Training J. Hodge, Sequoyah Projects Director A. Jenkins, Sequoyah Operations Director R. Joplin, TVA Corporate Nuclear Examination Program Manager M. Lovitt, Sequoyah Plant Support Director T. McMutuary, Sequoyah Chemistry Manager K. Michael, Sequoyah Maintenance Director
C. Reneau, Sequoyah Plant Manager
FACILITY POST-EXAMINATION COMMENTS AND NRC RESOLUTIONS
A complete text of the licensees post-examination comments can be found in ADAMS under
Accession Number MLXXXXXXXXX.
Item
RO Job Performance Measure A3, Determine Potential Total Dose for Valve Alignment
Comment from Facility Licensee
During the Initial License Operating Exam in 2022, the JPM developed for the Administrative
Topic Radiation Control it [sic] was determined to not be capable of completion. Specifically,
the valve stroke times were not located in the procedures referenced in the JPM.
The error in the JPM was discovered in the first round of administration of the JPM. The JPM
was administered to all RO candidates because the resolution was not immediately clear.
The Bank JPM was affected by a procedure change that removed the stroke times. The JPM
was not updated to the new procedure. During Validation an assumption was made to yield the
right answer. Based on different assumptions a few different stroke times could be found in
different procedures. Because the CUE statement was invalid and referred to stroking a valve
per a procedure that didnt contain guidance for that specific valve and the need for an
assumption to determine the correct time, the JPM was deemed INVALID.
A new similar JPM was developed, validated, approved, and administered to the RO
candidates.
NRC Resolution
The licensees recommendation was accepted.
The RO A3 administrative JPM that was approved to be administered before the operating test
weeks was designed so that applicants could use procedure 1-PI-OPS-000.003.0, Periodic
Stroking of Unit 1 Time Critical Valves to determine the maximum stroke times for proposed
test work on several valves. Once the maximum stroke times were determined, these times
would be used along with radiological maps to determine the expected dose that would be
received by operators performing the assigned task. This JPM was a bank JPM.
However, during the administration of this admin JPM to the RO applicants, the NRC examiners
determined that the maximum valve stroke times for the proposed valves had been removed
from the intended procedure. This unanticipated condition meant that the applicants had to
assume a maximum valve stroke time that could be determined to be widely divergent,
depending on the procedural reference and method that the various applicants chose to use (an
unstated assumption) to determine the maximum stroke time. In effect, the applicants were
forced to make unwarranted assumptions to determine maximum stroke time; this in turn
resulted in widely varying answers and an inability to evaluate the JPM in a consistent and
equitable manner.
The NRC examiners immediately contacted the Branch Chief, who happened to be onsite
during the initial days of the operating test administration for management observation
purposes, and obtained his permission to declare this JPM an invalid JPM in accordance with
the guidance of NUREG-1021 section ES-3.5 B(2), which states in part: substitute or
replace planned material only if an item is determined to be invalid or impossible to perform or
simulate Operating test changes require NRC Chief Examiner and regional Branch Chief
approval.
Another radiological controls administrative JPM was developed, validated, and ultimately
administered to all RO applicants following approval by the Chief Examiner and regional Branch
Chief.
________________________________________________________________________
Item
Scenario 5, Event 4: Trip of the 1A Charging Pump (CCP 1A)
Comment from Facility Licensee
During a simulator session, a candidate stated that in MODE-1 TS-3.5.5. Seal Injection Flow
was applicable. This was not listed in the Scenario Guide as a required condition.
The Scenario 5 guide did not list any applicable Tech Spec items for Event 4. This was
questioned by the NRC Examiner team and asked for [sic] Operations Department to review the
required Tech Specs for the given condition.
Operations Department determined the following conditions are applicable for event 4 and
should have been listed in the Scenario Guide:
TS-3.5.2, Condition A
TRM-8.1.1 Reactivity Control Systems, Condition A
NRC Resolution
The licensees recommendation was accepted.
During the administration of Scenario 5, the NRC examiners identified that the detailed scenario
guide did not contain the complete Technical Specifications (TS) applicable to the charging
pump trip that were needed to correctly evaluate the SRO applicants on the TS for this event.
The examiners requested that the facility licensee investigate and provide updated information
via post-exam comment, if changes to the scenario guide were needed.
The NRC agrees with the facility that the appropriate specifications for this event were to enter
TS LCO 3.5.2, Emergency Core Cooling Systems (ECCS) - Operating, CONDITION A; and
Technical Requirements Manual (TRM) LCO 8.1.1, Reactivity Control Systems, CONDITION
A. All applicants who were administered Scenario 5 were evaluated against these required
specifications for the Technical Specification competencies.
________________________________________________________________________
Item
Question 50, K/A 073G2.2.4
Comment from Applicant
There is confusion on what specific equipment to which the question is referring. 0-RE-090-
25-A is a UNID that is typically used to refer to the associated monitoring instrument and all
associated equipment for the Main Control Room Radiation Monitor. RE in the UNID
specifically refers to the radiation element within the radiation monitor, but it is widely accepted
to use that UNID for the entire radiation monitor including associated equipment as evidenced in
the annunciator response and drawings associated with the radiation monitor. To indicate the
confusion of the operator, the annunciator for high radiation from this radiation monitor (1-AR-
M12-B, window C-7) is displayed as 0-RA-90-125A MAIN CNTL RM INTAKE MON HIGH
RAD.
Here is a list of some UNIDs specific to the Main Control Room Intake Monitor:
0-RM-90-125 - Main Control Room intake monitor indicator
0-RM-90-125TA - Main Control Room intake monitor isolator
0-RE-90-125 - Main Control Room intake monitor radiation element
0-FI-90-125 - Main Control Room intake monitor flow indicator
0-RA-90-125 - Main Control Room intake monitor ratemeter
0-FS-90-125 - Main Control Room intake monitor flow switch
0-RI-90-125 - Main Control Room intake monitor rate indicator
0-PI-90-125 - Main Control Room intake monitor pressure indicator
As a challenge for the first part of the question, the applicants took the verbiage RE-90-125,
Control Room Ventilation Instrument to mean the equipment in its entirety as opposed to just
the indicating and alarm circuit. In this case, parts of that equipment (the indications) are
powered from Unit 1 while other parts of that equipment (sample pumps, temperature monitor)
are powered from Unit 2. With this information, Unit 1 or Unit 2 would be the correct answer
based on what part of the equipment you are discussing. The question, as written, is not
detailed enough in asking which part of the equipment it is asking about as a power supply (see
0-SO-90-2 Attachment 1).
For the second part of the question, the stem does not state only RE-90-125 in alarm, it
specifically states when RE-90-125 is in alarm. This implies that with all equipment operating
correctly (no instrument malfunction on any radiation monitor nor [sic] other abnormal condition
stated in the stem), the operators thought process would be reasonable to think that an actual
high radiation signal is what caused RE-90-125 to alarm to high. With a high radiation signal
causing RE-90-125 to alarm, and with no other equipment stated to be out of service or
malfunctioning in the stem, it is also reasonable to assume RE-90-125 (A-train) and RE-90-126
(B-train) would both actuate causing both trains to actuate their respective Control Room
Isolation signals. This is given, since they sample from the same location connected to outside
of the control building in order to redundantly protect the MCR envelope from high radiation (see
47W866-4).
Recommend changing the answer to B and D as correct answers.
Facility Licensee Recommendation
For the first part of the question, RE-90-125 has specific power supplies from both units for the
monitor and associated equipment. For the second part of the question, it is correct that no
statement was made in the stem for redundant equipment out of service or an abnormal
condition such as an instrument malfunction on RE-90-125 or RE-90-126. Therefore, any
legitimate (non-spurious) high radiation signal that actuates RE-90-125 would also actuate RE-
90-126. Considering this, both trains of control room isolation would actuate in this condition,
making B and D correct for the second portion of the question. As such, request the answer key
be modified to accept B and D as the two correct answers; or remove the question from the
exam as determined to be in accordance with ES-4.4 section C.3.c.
NRC Resolution
The licensees recommendation was rejected.
One of the contentions associated with this question was that the question was confusing
because there are associated components to radiation monitor RE-90-125 that are powered
from Unit 2 as well as Unit 1. However, there was no listed answer choice provided that stated
both Unit 1 and Unit 2, and there were no applicant questions asked concerning this question
during the written examination administration on May 18, 2022. The lack of this particular
distractor is, in and of itself, an indication to the applicants that the question is asking the power
supply to a particular/specific component, and not multiple components within the same system.
Of note, the first part question does not ask for the power supply to the pump, ratemeter, flow
indicator, or any other associated component to RE-90-125; it simply stated RE-90-125,
Control Room Ventilation Instrument, is powered from ___. It is technically correct that RE-90-
25 is powered via a breaker on Unit 1 120VAC Vital Instrument Board 1-I; therefore, the only
assessed correct answer to the first part question statement is Unit 1.
The region disagreed that the second part question statement directly concerned a valid high
radiation condition. The second part statement reads as follows: When RE-90-125 is in HIGH
Alarm, [Only one/two] train(s) of Control Room Isolation will automatically actuate. There are
no other conditions provided in the question stem that indicate a radiological accident is in
progress, or that there is an actual high radiation condition present at the common control room
intake that would cause any other radiation monitors to also be in HIGH Alarm. For example, it
is entirely possible that the RE-90-125 HIGH Alarm condition could have occurred during
normal/routine instrumentation and controls channel testing of the radiation monitor, and it
would be important for operators to understand that for channel testing of RE-90-125 only one
channel of control room isolation would actuate, not two. The contention that an applicant was
required to assume that an actual high radiation condition existed, which would also put RE-90-
26 in alarm and cause a total of two trains to automatically actuate, is an unwarranted
assumption that is prohibited by NUREG-1021 ES-1.2.
Therefore, in accordance with the requirements of NUREG-1021 ES-4.4, the regional examiners
determined that answer choice A remained the only technically correct answer to Question 50,
and no changes were made to the approved answer key for Question 50.
________________________________________________________________________
Item
Question 61, K/A 045A4.10
Comment from Applicant
[Note the following initial information:] References: 0-GO-4 R110
Definitions:
EHC - Electro-hydraulic Control System (previously installed analog system)
DEHC - Digital Electro-hydraulic Control System (current digital system)
This question locks the operator into making an absolute choice on whether the Steam
Generators will shrink or swell upon TV-GV transfer (throttle valve/governor valve transfer).
With the previously installed EHC system, the TV-GV transfer process commenced by the
throttle valve opening at 1700 RPM, then the governor valve throttling closed to reduce the
increased speed in the main turbine, known as catching the turbine close to 1800 RPM. This
operation is reflected by the caution in 0-GO-4. The screenshot below [not provided in this
summary] is from 0-GO-4 and a small excerpt is also included from the original Westinghouse
manual on EHC describing the TV-GV transfer process.
[reference omitted]
With the described process prior to DEHC upgrade, the expected response typically occurs, and
the SG level may swell, as indicated on the caution in 0-GO-4.
With the plant upgrade to DEHC, the system design was changed to eliminate the possibility of
swell of the SG levels and to maintain the positive control of turbine speed (reduced the
catching effect), see attached graphs indicating pre and post DEHC TV-GV transfers (M1 to M3
on attached graphs).
[reference omitted]
With the DEHC system installed, the governor valves (GVs) lower speed an incremental value
(10 RPM) to ensure positive control prior to the throttle valves (TVs) fully opening. The lowering
of turbine speed/steam demand will cause a negligible effect of steam generator level for two
reasons (1) the turbine is not under load so slowly lowering speed 10 RPM on a free spinning,
connected turbine shaft is less than.0025% of total steam flow and (2) that amount of steam
flow change would be split among the four steam generators to further reduce any change in
indicated steam generator water level due to shrink which would be negligible and near [sic]
impossible to detect by visual or graphical analysis. The entire TV-GV transfer process from
TVs controlling at 1700 RPM to GVs controlling at 1800 RPM takes several minutes by design,
as to prevent any perturbation to the system.
Recommend deleting the first position of this question making B and D the correct answers.
Facility Licensee Recommendation
The recent upgrades to the turbine control system would constitute newly discovered technical
information that supports a change in the answer key as described is ES-4.4 C. Since shrink
nor [sic] swell occurs anymore upon TV-GV transfer with the DEHC system (per design),
recommend changing the answer key to B and D being correct since the second part of the
question is not being challenged. Since B and D have mutually exclusive information in the
answer for part 1 (shrink and swell), removing the question from the exam may be determined
as appropriate per ES-4.4
NRC Resolution
The licensees recommendation was rejected.
No applicants asked questions concerning Question 61 during written exam administration on
May 18, 2022.
The NRC regional office agrees with the applicant and facility contentions that the recent plant
modifications involving the DEHC system are designed to minimize the effects of shrink or swell
in Steam Generator (S/G) levels during the TV-GV transfer operation. Furthermore, the
provided parameter trends show clearly that the actual plant response during TV-GV transfer
operation is neither a shrink nor a swell event in any appreciable terms that would be evident to
any operator.
However, the regional office notes that the specific language of the question reads as follows:
Which one of the following completes the statements below in accordance with 0-GO-4, Power
Ascension from less than 5% Reactor Power to 30% Reactor Power? When the TV-GV
transfer occurs SG level [swell or shrink] is expected. This questions which one of the
following statement clearly associates that the correct answer to the question is derived from
the procedure (0-GO-4), not the actual plant response. Furthermore, the first part question
statement asks what is the expected plant response per the procedure, not to describe the
actual plant trend response in the absence of operator actions.
On page 56, procedure 0-GO-4 contains a CAUTION statement that reads as follows: 1) SG
level may swell during TV-GV transfer. Moreover, step [43.2] reads as follows: NOTIFY S/G
level operator to anticipate a level swell. Step [43.2] is immediately followed by the step to
press the TV-GV TRANS button to perform the automatic TV-GV transfer. Procedure 0-GO-4
has been updated to account for the recent DEHC system, and yet still contains the CAUTION
and step [43.2] as stated above. Therefore, it is technically accurate that procedure 0-GO-4
directs the operators to anticipate a potential S/G level swell, even with the installed DEHC
system. To require the operators to anticipate a S/G swell effect, therefore, is completely
analogous to asking whether a shrink or swell is expected in accordance with 0-GO-4, which is
what the question statement actually says.
Per the above line of reasoning, the regional office determined that answer choice B remained
the only technically correct answer to Question 61; therefore, in accordance with NUREG-1021
section ES-4.4, no changes were made to the approved answer key for Question 61.
________________________________________________________________________
Item
Question 84, K/A 076G2.2.22 (SRO only)
Comment from Applicant
This question asks the student to determine if the load reduction/thermal power change is within
the rate of change limits of 0-GO-5 (Normal Power Operation). Based on the context given in
the question, it is assumed that the question intended to test the students knowledge on the
limitations of 0-GO-5 relating to power/reactivity changes. This analysis will look at the
applicable limitations of 0-GO-5 to determine if the down power was within or exceeded any
limits in 0-GO-5.
When referring to 0-GO-5, the stem specifically asks about limits. This implies that there are
several limits associated with power and rate of change within 0-GO-5. There are several rate
of change limits in 0-GO-5 (see section 3.2 Limitations in 0-GO-5):
-3.2[A.]: Do not exceed a load change rate of plus or minus 5% per minute or a step change of
10%. -MET
-3.2-[E.]: For each instance of change in the rated thermal power level exceeding 15% in one
hour, Chemistry must be notified to initiate the conditional portions of 0-SI-CEM-000-877.0, 0-
SI-CEM-030-407.2 and 0-SI-CEM-030-415.0 due to the thermal power change and ensure
narrative log entries are made for each power change exceeding 15% in one hour (SR 3.4.16.2)
- DID NOT MEET
The stem of the question gives the operator power levels in RTP (Rated Thermal Power
measured in MW Thermal) and not load (MW Electric or MWe) to analyze that a limit was or
was not exceeded per 0-GO-5. While the load reduction value (initial and final value in MWe)
was not given, the initial and final value of rated thermal power was given indicating the
reduction focuses on thermal power and not load or MWe.
A rate limitation that was exceeded in this question was the 0-GO-5.3.2 [E.] limit that requires
actions by chemistry when rated thermal power is reduced by 15% or greater in a one-hour
period. The stem then asks the chemistry specific actions in the second part of the question if
this limit (15% RTP or greater change in one hour) were [sic] to be exceeded.
Recommend changing the answer to A for this question since a limit in 0-GO-5 was exceeded.
No challenge on the second part of the question.
Facility Licensee Recommendation
A rate of change limit in 0-GO-5 was exceeded in the form of 15% power (RTP) change in less
than a one hour period. Since the questions [sic] asks if limits were exceeded, based on the
information provided, a rate of change limit of 0-GO-5 was exceeded (15% power reduction in
one hour) making A the only correct answer. Request the answer key be changed to reflect
A as the correct answer in accordance with ES-4.4 section C.3.c.
NRC Resolution
The licensees recommendation was partially accepted.
The NRC regional examiners did not agree with the applicant and the facility contention that the
requirement to notify the Chemistry department to perform certain Surveillances following a
specified amount of rated power change constituted a rate of change limit as specified by the
question stem. The first part question specifically asks this reduction [exceeded/was within]
the rate of change limits. Just because the Chemistry SR information was also contained in 0-
GO-5 section 3. Limitations, does not mean that all of the information in this section was
related to a rate of change limits.
However, the regional examiners are required to analyze the complete question and make the
most appropriate technical determination concerning the question contents. In this case, the
question stem only informed the applicants that an unplanned load reduction was started at time
0000, and that by time 0015 the load reduction was completed to 80% RTP. The question
authors intended to exercise the rate-of-change limits stated in 0-GO-5 3.2[A], which stated: Do
not exceed a load change rate of plus or minus 5% per minute or a step change of 10%.
During question development and validation, it was assumed that the given power change was
made in a constant and approximately linear fashion, so that the applicant would correctly
determine a 20% power change in 15 minutes or 1.33% power change per minute, which is
within the 5% per minute limit stated above.
Unfortunately, the question did not specify a power profile, or state that the unplanned power
change was made at a constant or linear rate. Therefore, in a practical technical sense, what
this meant was that any number of power reduction profiles could have been performed by the
operators. Although a constant/linear rate of change would result in a determination of within
limits, the plant power could have been changed, for example, 12% in the first minute of the
power reduction and then 8% change in the ensuing 14 minuteswhich would have resulted in
a violation of the rate-of-change limits. The applicants were not provided with a power profile or
any other information that would allow them to determine how the plant lowered power for this
questions purposes, without making an unwarranted assumption that may have resulted in
multiple valid/technically correct answers for the first part question. The regional examiners
accordingly determined that the first part question did not provide the applicants enough
information to correctly answer the question.
There were no contentions made against the second part question, and the regional examiners
could not determine any issue with the technical information and correct answer to the second
part question. Therefore, in accordance with NUREG-1021 Section 4.4, the correct answer for
Question 84 was changed to both A and C, and all SRO applicants were graded in
accordance with the modified answer key.
________________________________________________________________________
Item
Question 90, K/A 073G2.4.41 (SRO-only exam)
Comment from Applicant
The operator was given the knowledge that the unit has a leaking fuel assembly (high RCS
activity) and that the unit also has a tube leak (stem informed the operator this happened at
0745). In this instance, the NOUE criteria was met just prior to 0800 and the ALERT criteria
was met at some time prior to 0815.
See EPIP-1, an ALERT declaration requires 1-RM-90-256 to be above a reading of 11,200
mR/hr for 15 minutes or longer. At 0815, 1-RM-90-256 reading was 16,400 mR/hr which
indicates that the ALERT criterion for that radiation monitor was exceeded some time before
0815. With 60 minutes given to declare the NOUE and 15 minutes given to declare the ALERT,
no correct answer is given from which to choose. The monitor indicated an ALERT value prior
to 0815, so the answer could not be 0830 as you would have exceeded 15 minutes prior to this
time.
Recommend removing the questions from the exam since there is no correct answer and it is
unreasonable to determine when the ALERT criterion was exceeded (to the minute) without a
plotted graph.
Facility Licensee Recommendation
Based on the information given, it is accepted to reasonably determine that at some point in
time before 0815 the ALERT criteria was met. In this instance, no correct answer exists for the
+15 minute requirement to declare an ALER
- T. In an actual event, the ALERT declaration would
be late if it was declared at 0830. There was no note in the stem stating to not interpolate data
or to only view the data timeframes given which confused the operators. Per ES-4.4 section C,
recommend removal of this question from the exam for a confusing or misleading stem to where
there is no correct answer.
NRC Resolution
The licensees recommendation was accepted.
No applicant questions were asked concerning Question 90 during written exam administration
on May 18, 2022.
NRC regional examiners agreed with the applicant and facility licensee contention that there
was no technically accurate answer to Question 90. During question development, the intention
was that the applicants would use the provided time data without having to interpolate or make
assumptions regarding the actual times that the various thresholds for a Notice of Unusual
Event (NOUE) or ALERT EAL were reached. The question statement language of Based on
Radiation Monitor Readings, was intended to convey this point. However, upon further
reflection, the examiners agreed that the question stem and question statement were not clear
enough to alleviate any confusion.
Moreover, it is more technically precise/accurate to determine that the NOUE and ALERT
thresholds were reached at some indeterminate time that was not specified to the applicants.
For example, the questions provided data (at 15-minute intervals) did not include additional
data points giving the precise time that the threshold radiation monitor readings were reached.
Also, the question did not direct the applicants to use linear radiation monitor trends to calculate
the precise times that the threshold radiation monitor readings were reached. Determination of
the precise time the thresholds were reached was a required aspect to correctly answering the
provided question (e.g., the actual correct answer to the question was 15 minutes after the first
EAL threshold was reached).
Accordingly, the regional examiners determined that there was no technically correct answer to
this question. NUREG-1021 section ES-4.4 requires that a question that has no technically
correct answer to be deleted from the examination; therefore, in accordance with NUREG-1021,
the examiners deleted Question 90 from the examination in its entirety, and all SRO applicants
were graded against 24 SRO-only questions and 99 questions for the overall score.
SIMULATOR FIDELITY REPORT
Facility Licensee: Sequoyah Nuclear Plant
Facility Docket No.: 05000327 and 05000328
Operating Test Administered: May 2 - 11, 2022
This form is to be used only to report observations. These observations do not constitute audit
or inspection findings and, without further verification and review 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.
While conducting the simulator portion of the operating test, examiners observed the following:
Item
Description
During NRC Exam conducted on 05/02/2022 RCP seal failure
occurred prior to reactor scram. When seal failure variables
cval1damage() increased, the first four shutdown bank A rod
positions fluctuated between 228 and 0. This resulted in an
unexpected power drop.
During the NRC exam on 05/04/2022 malfunction CC05C
resulted in the Unit 2 CCS Surge Tank level dropping to the point
of makeup initiation sooner than the Unit 1 CCS Surge Tank.
Also, the rate of makeup seemed to be larger than expected.
CC05C leak rate was found to be 194 GPM.