IR 05000335/2020301

NRC Operator License Examination Report 05000335/2020301 and 05000389/2020301

ML21047A496
Person / Time
Site:	Saint Lucie
Issue date:	02/16/2021
From:	Gerald Mccoy NRC/RGN-II/DRS/EB1
To:	Moul D Florida Power & Light Co
References
Download: ML21047A496 (21)

Text

February 16, 2021

SUBJECT:

ST. LUCIE NUCLEAR PLANT - NRC OPERATOR LICENSE EXAMINATION REPORT 05000335/2020301 AND 05000389/2020301

Dear Mr. Moul:

During the period December 7 - 12, 2020, the Nuclear Regulatory Commission (NRC)

administered operating tests to employees of your company who had applied for licenses to operate the St. Lucie Nuclear Plant. At the conclusion of the tests, the examiners discussed preliminary findings related to the operating tests and the written examination submittal with those members of your staff identified in the enclosed report. The written examination was administered by your staff on December 17, 2020.

Two Reactor Operator (RO) and six Senior Reactor Operator (SRO) applicants passed both the operating test and written examination. One SRO applicant received an excusal for the operating test and passed the written examination. One SRO applicant passed the operating test, but did not pass the SRO-only portion of the written examination. There were five post-exam comments submitted on the written exam and two post-administration comments concerning the operating test. These comments, and the NRC resolution of these comments, are summarized in Enclosure 2.

The initial operating test examination submitted by your staff did not meet the guidelines for quality contained in NUREG-1021, Operator Licensing Examination Standards for Power Reactors, Revision 11, as described in the enclosed report. The RO and SRO written examinations did meet the NUREG-1021 quality guidelines.

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) 997-4551.

Sincerely,

/RA/

Gerald J. McCoy, Chief Operations Branch 1 Division of Reactor Safety Docket Nos: 50-335, 50-389 License Nos: DPR-67, NPF-16

Enclosures:

1. Report Details 2. Facility Comments and NRC Resolution 3. Simulator Fidelity Report

REGION II==

Examination Report Docket No.: 05000335, 05000389 License No.: DPR-67, NPF-16 Report No.: 05000335/2020301, 05000389/2020301 EPID No.: L-2020-OLL-0022 Licensee: Florida Power and Light Company (FP&L)

Facility: St. Lucie Nuclear Plant, Units 1 & 2 Location: 6351 S. Ocean Drive Jensen Beach, FL 34957 Dates: Operating Test - December 7-12, 2020 Written Examination - December 17, 2020 Examiners: M. Meeks, Chief Examiner, Senior Operations Engineer K. Kirchbaum, Operations Engineer, Chief Examiner Under Instruction J. Bundy, Operations Engineer T. Morrissey, Senior Resident Inspector (Examiner Qualified)

Approved by: Gerald J. McCoy, Chief Operations Branch 1 Division of Reactor Safety Enclosure 1

SUMMARY

ER 05000335/2020301, 05000389/2020301; operating test December 7-12, 2020 & written exam December 17, 2020; St. Lucie Nuclear Plant; Operator License Examinations.

Nuclear Regulatory Commission (NRC) examiners conducted an initial examination in accordance with the guidelines in Revision 11, 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.

The NRC developed the written examination outline. Members of the St. Lucie Nuclear Plant staff developed both the operating tests and the written examination. The initial operating test submittal did not meet the quality guidelines contained in NUREG-1021.

The NRC administered the operating tests during the period December 7-12, 2020. Members of the St. Lucie Nuclear Plant training staff administered the written examination on December 17, 2020. Two Reactor Operator (RO) and six Senior Reactor Operator (SRO) applicants passed both the operating test and written examination. One SRO applicant received an excusal for the operating test and passed the written examination. One SRO applicant passed the operating test, but did not pass the SRO-only portion of the written examination. Nine applicants were issued licenses commensurate with the level of examination administered.

There were seven post-examination comments.

No findings were identified.

REPORT DETAILS

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 the 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 of December 7-12, 2020.

The NRC examiners evaluated two Reactor Operator (RO) and seven Senior Reactor Operator (SRO) applicants using the guidelines contained in NUREG-1021. Members of the St. Lucie Nuclear Plant training staff administered the written examination on December 17, 2020. Evaluations of applicants and reviews of associated documentation were performed to determine if the applicants, who applied for licenses to operate the St. Lucie 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 St.

Lucie 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 11, 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 that the licensees operating examination submittal was outside the range of acceptable quality specified by NUREG-1021. The initial operating test submittal required multiple JPM replacements after the onsite validation for JPMs that did not meeting the standards of NUREG-1021. Also, multiple simulator scenarios were required to be modified during onsite validation week and re-validated during that week.

The initial events of one simulator scenario required re-validation during exam week prior to administration. Future examination submittals need to incorporate lessons learned.

The NRC determined that the licensees initial written examination submittal (RO and SRO) was within the range of acceptability expected for a proposed examination.

Two Reactor Operator (RO) and six Senior Reactor Operator (SRO) applicants passed both the operating test and written examination. One SRO applicant received an excusal for the operating test and passed the written examination. One SRO applicant passed the operating test, but did not pass the written examination. Nine applicants were issued licenses commensurate with the level of examination administered.

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 five post-examination comments concerning the written examination. A full copy of the licensee post-examination comments may be accessed in the ADAMS system under ADAMS Accession Number ML21034A531. A copy of the final written examination and answer key, with all changes incorporated, may be accessed not earlier than December 19, 2022, in the ADAMS system as ML21034A519.

4OA6 Meetings, Including Exit

Exit Meeting Summary

On December 15, 2020, the NRC examination team discussed generic issues associated with the operating test with Mr. Dan DeBoer, Site Vice President, and members of the St. Lucie Nuclear Plant staff via video conference. The examiners asked the licensee if any of the examination material was proprietary. No proprietary information was identified.

On February 2, 2021, the NRC examiners discussed the examination results, licensing actions, and other items to be documented in the examination report with Mr. Seth Duston, Site Training Manager, and other members of the St. Lucie Training department staff via video conference.

KEY POINTS OF CONTACT Licensee personnel B. Beltz, Safety Assurance and Learning D. DeBoer, Site Vice President S. Duston, Training Manager T. Fisher, Operations Instructor W. Godes, Licensing Manager C. Hill, Corporate Training Manager B. Hinze, Operations Training Supervisor K. Keith, Operations Class Mentor C. Martin, Chemistry/Radiation Protection Manager T. Ouret, Operations Training Supervisor K. Paez, Licensing T. Spillman, Assistant Operations Manager R. Story, Outage Manager R. Virgin, Operations Training Supervisor S. Wylie, Examination Author NRC personnel T. Morrissey, Senior Resident Inspector

FACILITY POST-EXAMINATION COMMENTS AND NRC RESOLUTIONS

The facility submitted seven post-examination comments as indicated in a letter which can be

found in ADAMS under Accession Number ML21034A531.

Post-Examination Comment #1: Question 12, K/A 056 AK3.01, Loss of Offsite Power /

Knowledge of the reasons for the following responses as they apply to the Loss of Offsite

Power: Order and time to initiation of power for the load sequencer

Comment:

Assumption was made that the timeline was too vague to determine when the Emergency

Diesel Generator (EDG) output breaker was closed. 4 [sic] applicants assumed that the EDG

output breaker was closed in the 10 seconds between 00:05:00 and 00:05:10 and not at time

00:05:10 as the author intended. Therefore, the HPSI pump would have started by design on its

second block prior to 00:05:10. Therefore, accept answers A and C as both correct.

Facility Licensee Recommendation:

The station does not have plant data that would support the EDG Output breaker closing any

sooner than 00:05:08, therefore, at no time in the provided information would the HPSI Breakers

be closed. The station recommends no changes to question 12.

NRC Resolution:

Although the NRC did not accept the comment or the facility licensee recommendation, the

answer for question 12 was changed from distractor C only to A only.

During the review of the technical accuracy of this question post-exam-administration, the NRC

determined that the examination key for the exam, as administered, did not identify the correct

answer. The question presented a timeline that stated the following information:

(1) following a Loss of Coolant Accident (LOCA) from 100% power at time 00:00:00, a

Safety Injection Actuation (SIAS) signal actuated at time 00:03:00;

(2) at time 00:05:00, a Loss of Offsite Power (LOOP) occurred; and

(3) at time 00:05:10, both the 1A and 1B EDGs had started and output breakers were

CLOSED.

The first part of the question required the applicant to determine whether or not the 1A/1B HPSI

pumps were running at time 00:05:10. The initial answer key, as submitted by the facility and

approved by the NRC, was that the 1A/1B HPSI pumps would NOT be running at time 00:05:10.

All parties agree that the HPSI Pumps receive an automatic start signal on a 6 second load

block after the associated EDG breaker closes. It typically takes 8 to 10 seconds for an EDG to

come to rated speed and voltage before the output breaker closes and loads begin to sequence

on to the ED

G. During a normal LOOP with a simultaneous SIAS signal present, it would be

expected that the HPSI pump would start 6 seconds following the EDG breaker closure. The

original answer to the question assumed that at 00:05:10, the EDG breaker had just closed and

it would be an additional 6 seconds (i.e., time 00:05:16) before the HPSI pumps would start.

The question, as written, stated that the SIAS signal occurred 2 minutes prior to the LOOP. The

SIAS actuation will cause the EDGs to receive an automatic start signal, independent of

whether or not a LOOP is present. Given the timeline presented by the question, the EDGs

started and would be running at rated speed and voltage with the respective output breaker

open when the LOOP occurs. In other words, at time 00:05:00, when the LOOP occurs, the

EDGs did not require 10 seconds to come to rated speed. Since they are already running and

ready to supply power to the deenergized busses, the EDG output breaker would close in

approximately 2 seconds post-LOO

P.

The NRC identified that St. Lucie Training MaterialProperty "Contact" (as page type) with input value "P.</br></br>The NRC identified that St. Lucie Training Material" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., PSL OPS 0711501, Emergency Diesel

Generators (rev 30), provided the following information:

SIAS followed by delayed LOOP: The SIAS will immediately auto start all

SIAS designated loads and start the EDG. When the UV/DV (LOOP) signal is

sensed, the bus loads are shed, and after a ~ 2 second time delay, the EDG

Output breaker will close and SIAS loads will sequence on. The reason for the

time delay was to ensure that pump motor residual fields would collapse

without causing a voltage/current spike on motor reload for those pump motor

breakers that do not load shed.

Therefore, based on the available technical information, the correct timeline would read as

follows:

(1) at time 00:03:00, SIAS actuation and auto-start of the 1A and 1B EDGs;

(2) at time ~00:03:10, 1A and 1B EDGs at rated speed and voltage with output breakers

open;

(3) at time 00:05:00, a LOOP occurred causing bus load shed;

(4) at time ~00:05:02, 1A and 1B EDG output breakers close and SIAS loads begin to

sequence on;

(5) at time ~00:05:08, 1A and 1B HPSI pumps automatically start due to +6 second SIAS

load sequencer

(6) at time 00:05:10, 1A and 1B HPSI pumps are running due to load sequence

Section D.1.b of ES-403 of NUREG-1021 stated the following, in part: newly discovered

technical information that supports a change in the answer key would be one case where it is

most likely to result in post-examination changes agreeable to the NRC. In accordance with

NUREG-1021, the newly discovered technical information supports the NRC determination that

the HPSI Pumps would be running by the 00:05:10 time mark, based on the timeline given in

the question; therefore, the correct answer was changed from distractor C only to A only. All

applicants were evaluated against this change to the written examination answer key.

Post-Examination Comment #2: Question 15, K/A 062 AA1.07, Loss of Nuclear Service

Water: Ability to operate and / or monitor the following as they apply to the Loss of Nuclear

Service Water (SWS): Flow rates to the components and systems that are serviced by the

SWS; interactions among the components

Comment:

The applicants understand that the Circ Water seal water system is designed to handle a loss of

Intake Cooling Water (ICW) with backup from the domestic water system, it is their opinion that

the Component Cooling Water (CCW) seal system would still see a small temperature change.

Therefore, accept answers C and D as both correct.

Facility Licensee Recommendation:

After receiving the feedback from the applicants and discussing system design and operation

with operations personnel, it is possible that the system does see some perturbation and

temperature change on the Circulating Water Pump Seal coolers. Therefore, the station

recommends accepting C and D as correct answers.

[N.B.: Following the initial receipt of the post-exam comments, NRC management requested

that the facility licensee be contacted to ascertain if additional technical information could be

provided on the backup cooling system to the Circulating Water Pump seal coolers. The

following response was obtained.]

Amplifying Information Provided Post NRC Feedback:

Given the different operating temperatures and pressures between Intake Cooling Water and

Domestic Water, and the basic manually-controlled throttling mechanism of the Circulating

Water Pump seal water flow; it is reasonable to assume that after SIAS, when Circulating Water

Pump seal water swaps over to the Domestic Water backup supply, the Circulating Water Pump

seal water flow and temperature will change, thereby causing a change to the packing gland

temperature.

The lubricating water for the pump seals and bearings comes from the Intake Cooling Water

(ICW) system through a series of self-cleaning and manually cleaned strainers. Flow is

monitored via a flow indicating switch at each pump and is controlled via a manually throttled

ball valve. An alternate supply of lube water comes from the domestic water system, which

draws suction from the City Water Storage Tanks (CWST).

During a SIAS, ICW supply to the Circulating Water Pumps will be isolated by the automatic

closure of MV-21-2/-3, A/B ICW TRAIN TO TCW HXS. A low-pressure condition created from

isolation of ICW will cause PCV-21-26, B/U LUBE WATER TO CW PUMPS, to OPEN and

supply backup cooling water to the Circulating Water Pump seals. This backup cooling water

will be of a different pressure (75 psig vs 40 psig) and temperature (74°F vs 67-88°F) than the

Intake Cooling Water system, and thus change the flow and temperature of water supplying the

Circulating Water Pump seals.

Additionally, 2-NOP-21.02, Circulating Water System Operations, contains instructions for

aligning alternate sources of lube water to the Circulating Water Pumps. Given a potential

system pressure difference between the alternate vs normal seal water source, the procedure

directs throttling of seal water flow to the 6-10 gpm band and validating adequate lube water

leak off from the Circulating Water Pump packing gland. During alignment of the alternate water

sources, changing seal water flow may require monitoring of packing gland temperature to

validate no rising temperatures. Based on these instructions, when aligning alternate sources of

lube water to the Circulating Water Pumps, some seal water temperatures changes are

expected.

PSL Station Recommendation: Therefore, the station still recommends accepting C and D as

correct answers.

NRC Resolution

The licensees recommendation was partially accepted; however, the NRC determined that the

question should be deleted from the written examination.

The question as written presented the applicants a condition where Intake Cooling Water (ICW)

was taken away from non-essential loads. The question then asked, between the Circulating

Water (CW) Pump Seals and the Instrument Air Compressors (IAC), which system would

experience a temperature change due to the ICW system alignment. The CW Pump seals,

which are normally directly cooled from ICW, will automatically shift to a Domestic Water cooling

source on loss of ICW pressure. The IACs would lose the heat sink entirely with a loss of ICW

cooling. The second part question statement required the applicants to determine if the CW

Pump seal coolers or the IACs [would] experience a temperature change due to the ICW

alignment. The initial answer key proposed by the facility and approved by the NRC was that

the IACs will experience a temperature change due to the ICW alignment, and that the CW

Pump seal coolers will not experience a temperature change due to the ICW alignment.

The station provided the NRC with temperature trends for ICW and Domestic Water. However,

these temperature trends for ICW and Domestic Water were obtained using portable

instrumentation, because there is no installed instrumentation at the CW pump seals to

determine actual temperature. Accordingly, the station was unable to provide direct plant data if

CW pump seal temperature changes based on the cooling source changing from ICW to

Domestic Water. Moreover, due to potential differences in cooling water temperatures and flow

rates from the two systems, it cannot be determined if the CW Pump seals will or will not have a

change in temperature when shifting to the backup cooling source. Note that the facility

recommendation quoted above was not technically specific concerning the temperature trend:

it is reasonable to assume that after SIAS, when Circulating Water Pump seal water swaps

over to the Domestic Water backup supply, the Circulating Water Pump seal water flow and

temperature will change, thereby causing a change to the packing gland temperature.

[emphasis added]

Given the benefit of hindsight, the question would have been better written to elicit which system

does, or does not, have an automatic cooling water supplied given the conditions of the

question stem. It could be argued that the lack of installed station instrumentation for CW pump

seal temperatures caused the question asked on the exam (which system experience[s] a

temperature change) to be non-operationally valid for inclusion on an initial licensed operator

examination.

Section D.1.b of ES-403 of NUREG-1021 stated the following, in part: a question with an

unclear stem that did not provide all the necessary information would be one example of a

question most likely to result in post-examination changes agreeable to the NRC.

Furthermore, the same section also stated: If there is no correct answer, the question shall

be deleted.

Based on the information provided by the facility licensee and the NRC evaluation, the NRC

determined that not enough given information was provided in the question stem for an

applicant to determine a correct answer; in fact, it was indeterminate that such an answer could

be determined at all based upon a lack of installed plant equipment. In accordance with

NUREG-1021, the NRC decided that in the lack of a definite correct answer, the question

should be removed from the written examination. All applicants were evaluated against this

change to the written examination answer key.

Post-Examination Comment #3: Question 21, K/A 032 AA2.07 Loss of Source Range NI /

Ability to determine and interpret the following as they apply to the Loss of Source Range

Nuclear Instrumentation: Maximum allowable channel disagreement

Comment

The question stem states that the Reactor has been determined to be critical. However,

the applicants identified that the both of the Wide Range Nuclear Instruments (WRNI)

indications were approximately TWO decades too low for where the Reactor would normally be

while criticality is declared (normally near 1 x 10-5 %). Additionally, the information provided in

the question only shows WR power rising ~ ONE decade (from 3 x 10-8 to 3 x 10-7). This is also

contrary to normal reactor behavior during an approach to criticality, where WRNI power

typically rises by approximately two decades. Actual Unit 2 plant data from March of 2020

during a reactor startup shows that the Reactor went critical at a WRNI power of 1.89 x 10-5

following a rise of over two decades. This expectation has also been validated during

performance of 2-PTP-81, Reload Startup Physics Testing, which showed WRNI power (after

raising power two decades above critical) was recorded as 1.0 e-3, so criticality likely occurred

at ~1.0 e-5.

Therefore, the response of the WR nuclear instruments would indicate that the reactor should

only be half way to criticality. This would align with the response of S/U channel B (which has

only shown 3 doublings). That means the response of 3 out of 4 Nuclear Instruments show the

reactor only being half way to criticality, leading the applicants to determine that Startup

Channel A is NOT reading correctly, and Startup Channel B IS reading correctly. Therefore,

accept answers A and C as both correct.

Facility Licensee Recommendation:

The question provided the candidates with 2 sets of nuclear instrument data for the candidate to

determine which set of data was expected for the conditions. The correct answer is based on an

original source range value that then doubles 6 times therefore by definition, the reactor is

critical. However in March of 2020, the most recent Unit 2 Startup, the Reactor went critical at

1.89 x 10-5 with nuclear instrument response on par with the candidates description. Due to the

new information related to the March 2020 reactor startup, the station is recommending

accepting A and C as correct answers.

NRC Resolution

The licensees recommendation was not accepted, and no changes were made to the written

examination answer key.

The licensee recommends that both A and C are correct answers. However, it is provided in the

question that, Below is the listed initial power levels power to commence approach to criticality

and the power when the reactor was announced critical

This is a given statement in the question, and cannot be assumed to be an incorrect statement

within the bounds of the question. In other words, based on the given information, the applicant

must analyze the question keeping in mind that the reactor actually is critical.

It is expected that reactor power will double 5 to 7 times (5-7 doublings) from the beginning of

the reactor start up to the point that the reactor is critical, as indicated on Source Range

instruments. The final reactor power at the point of criticality is dependent on the source term,

or source neutrons, present at the beginning of the reactor start up. This value can vary

significantly from one reactor startup to another. The longer a reactor is shutdown, the lower

the initial source term will be, based in part upon the decay of the various neutron emitting

sources within the reactor core. Yet there are still 5-7 doublings of the source term, which is

proportional to source range power level during a reactor start up to Criticality. The point at

which a reactor can go critical can be well below the Wide Range Power indication lower scale,

given the time since shutdown has been significant.

The licensee position is that distractor choice C, which corresponds to SR Channel B ONLY

indicating as expected, is also a correct answer. But SR Channel B indication went only from

cps to 400 cps which corresponds to 3 doublings. This is outside the expected range for

criticality, and since it is given that the reactor is critical, this indication must be incorrect or

suspected as being inoperable. And since it was provided that the reactor was critical, this

cannot be a correct answer.

The Wide Range indication provided in the question is low as compared to the single, recent

start up that is provided by the licensee for reference. However, as stated in 2-GOP-302,

Precaution 2.1.1, Criticality shall be anticipated whenever positive reactivity additions are being

made (i.e., CEA withdrawal, boron dilution, etc.). Although the provided indications are lower

than those observed during the last core start up, there are multiple variables that can affect the

WR power level at critical conditions. An example of a critical reactor with an extremely low or

off-scale WR value can be found during startups following extremely long shutdowns where the

source term is extremely small. The 5-7 doubling to Criticality applies, even when WR are

barely on scale.

Based on the above discussion, the NRC determined that the question as written was

technically accurate and the question had only one correct answer. No changes were made to

the answer key for question 21.

Post-Examination Comment #4: Question 32, K/A 006 K6.13 Emergency Core Cooling /

Knowledge of the effect of a loss or malfunction on the following will have on the ECCS: Pumps

Comment:

The question stem doesnt contain enough information to determine if a HPSI pump restart

would be allowed. One attempt to restart the HPSI pump would be allowed to protect the health

and safety of the public. Since the stem provides no mention of this and it is not the ROs

function to determine if health and safety of the public is at risk, and there is no other

information regarding why the HPSI pump tripped, it is rational that starting the HPSI pump one

time would be allowed. Therefore, accept answers A and B as both correct.

Facility Licensee Recommendation:

The stem of the question does not provide enough information for the RO candidate to

determine the impact on the HPSI system and associated pump. make [sic] the decision based

on health and safety of the public. Therefore, the station recommends accepting A and B as

correct answers.

NRC Resolution:

The licensees recommendation was not accepted, and no changes were made to the written

examination answer key.

The licensee provided insufficient information for the basis of the comments. Safety related

pumps and breakers can be restarted/reclosed following a trip, provided there is no indication

that there is damage to the pump, system, or electrical components. The deterministic portion

of that requirement is that the restart is to protect the Health and Safety of the Public. To

determine that the plant conditions provided would lead to a direct threat to the Health and

Safety of the Public required assumptions to be made outside of the conditions provided within

the question. The statement that this is higher than RO level knowledge was not substantiated

by the station with any procedural requirements, job description, or specific training that would

support this statement.

Provided in the stem of the question was enough information that the applicant was to

determine that 2-EOP-99, Figure 2, Safety Injection Flow vs. RCS Pressure, was being met for

Single Full Train operation. Based on this information is to be expected that a licensed operator

would recognize that this meets the minimum required safety functions per 2-EOP-03, LOC

A.

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Public. Therefore, any assumption that there was a threat to the Health and Safety of the public

would be false.

Based on the above discussion, the NRC determined that the question as written was

technically accurate and the question had only one correct answer. No changes were made to

the answer key for question 32.

Post-Examination Comment #5: Question 99, K/A G2.4.29, Knowledge of the emergency

plan.

Comment:

Based on the initial conditions in the stem of the question, a loss of Offsite Power did not occur.

I know this based on SBCS being used for cooldown; therefore, the main condenser is available

with the circulating water pumps running. Since there was no loss of offsite power, main

feedwater would have been feeding both S/Gs post trip through the low power Feedwater

regulating valve restoring S/G water level towards the normal program band of 60%-70%.

Then, the questions stem states that, AFAS-1 and AFAS-2 actuated. This would only occur if

S/G water level in both the A and B S/G reached a level of 19% NR for a total of 210

seconds. The only way for S/G water levels to reach the value for AFAS-1 and AFAS-2

actuation, would have been as a result of losing main feedwater. Due to the SGTR being in

progress, a SIAS must have occurred. When SIAS occurs, main feedwater pumps will be lost

due to receiving a trip signal from SIA

S. At the time Main Feedwater is lost, S/G water level will

be above 19% NR (AFAS-1 & AFAS-2 actuation setpoint).

The RCS Heat removal safety function per 2-EOP-04 is as follows [picture of 2-EOP-04 p. 58

included]:

The RCS heat removal safety function requires that either, an unisolated S/G level is between

60%-70% NR with Tcold stable or lowering, or feedwater is being controlled to restore the

unisolated S/G level to between 60%-70%, with Tcold stable or lowering. Once Main Feedwater

is lost, the RCS heat removal safety function is not being met, and immediate action is required

to restore the RCS safety function.

ADM 11.16 specifically states: If the safety function status check acceptance criteria are NOT

met for a particular safety function, the operator should immediately report this to the Unit

Supervisor, then the operating crew should take appropriate contingency actions necessary to

restore the safety function.

Depending on how quickly the crew starts AFW (Auxiliary Feedwater) and what S/G water

levels are at, they may have to limit themselves to 150gpm for 5 minutes per the hard card.

This could cause S/G levels to continue to lower for the first five (5) minutes, due to the high

steaming rate in effect to cooldown the RCS to less than five hundred ten (510) degrees F for

S/G isolation. The lowering level could result in S/G levels going below 19% narrow range for

sufficient time to allow AFAS to time out and actuate, as stated in the stem, even after AFW is

manually started by the crew.

The desk RCO will restore S/G water level IAW 2-NOP-99.07 Operations Hard Cards

5: the hard card has the operator close the steam admission valve to the C AFW

Pump from the ruptured S/G [picture of 2-NOP-99.07 p. 18, 19, and 20 included].

Due to restoring S/G water level IAW with the operations hard card, the steam admission valve

from the affected S/G to the C AFW pump would be closed by the time AFAS 1 and AFAS 2

would have actuated. Although, not starting the 2C AFW pump, step 3A is performed by the

operator to prevent the steam admission valve from automatically opening on a subsequent

AFA

[[contact::S. [sic] Actuation. Both motor driven AFW pumps are available]], therefore, there is no

reason to use the 2C AFW pump. By steaming the ruptured S/G to the main condenser and

with the steam admission valve to the C AFW pump being closed, no release is currently in

progress.

Additionally, ADM-11.16 [CAUTION statement] states: IF Main Feedwater is lost and AFAS is

left to automatically actuate, damage to the S/G feed ring will occur. Prompt restoration of

Steam Generator levels using Auxiliary Feedwater is allowed in order to satisfy the Safety

Function prior to formally addressing the entire RCS Hear removal Safety Function. The allows

the crew to restore Steam generator water level with Auxiliary Feedwater.

IAW EPIP-02 a release is defined as follows: A Steam Generator, with primary-to-

secondary leakage, due to a tube leak or rupture, is vented to atmosphere; Operating the C

AFW pump with steam being supplied from a Steam Generator with primary to secondary

leakage due to a tube leak or rupture. Then a RELEASE is in progress. The ruptured S/G is

not being vented to atmosphere, and the C AFW pump steam admission valve from the

ruptured S/G would have already been closed in accordance with Plant procedures. Therefore,

a release is NOT in progress.

Facility Licensee Recommendation:

PSL Station Recommendation: Based on the comment submitted by the Applicant, the Station

agrees with the applicants statement that due to the information provided in the steam of

question 99, the applicant, while following the guidance provided to the applicant during the

NUREG- 1021, Appendix E briefing, which states: If you have any questions concerning the

intent or the initial conditions of a question, do not hesitate to ask them before answering the

question. Note that questions asked during the examination are taken into consideration during

the grading process and when reviewing requests for informal NRC staff reviews (appeals). Ask

questions of the NRC examiner or the designated facility instructor only. A dictionary is available

if you need it.

When answering a question, do not make assumptions regarding conditions that are not

specified in the question unless they occur as a consequence of other conditions that are stated

in the question. For example, you should not assume that any alarm has activated unless the

question so states or the alarm is expected to activate as a result of the conditions that are

stated in the question. Similarly, you should assume that no operator actions have been taken,

unless the stem of the question or the answer choices specifically state otherwise. Finally,

answer all questions based on actual plant operation, procedures, and references. If you believe

that the answer would be different based on simulator operation or training references, you

should answer the question based on the actual plant.

The assumptions made by the candidate are consistent with the guidance provided.

Therefore, the station recommends accepting A and C as correct answers.

NRC Resolution:

The licensees recommendation was not accepted; no change was made to the answer key for

question 99.

The question as written provided the following given information in bulleted format, not

timeframe format:

(1) Unit 2 is experiencing a SGTR

(2) 2-EOP-04 SGTR is in progress

(3) Initial Cooldown for S/G isolation is in progress using SBCS

(4) AFAS-1 and AFAS-2 have actuated

Then, given the above information, the first part question statement asked the applicant to

evaluate if IAW EPIP-02, Duties and Responsibilities of The Emergency Coordinator, a release

(is or is NOT) currently in progress. The contention therefore resolves itself into two different

determinations: (1) what is meant by the term currently in the question statement; and then (2)

based on what currently refers to, is the (steam-driven) 2C AFW pump being operated while

supplied from a Steam Generator with primary to secondary leakage due to a tube leak or

rupture. As correctly noted by the applicant contention quoting 2-EPIP-02 above, if the 2C AFW

pump is running and yet-to-be-isolated from the ruptured S/G at the currently point of the

question statement, a release is in progress; if the 2C AFW pump is not running or has been

isolated from the ruptured S/G at the currently point of the question statement, a release is

NOT in progress. We now examine these points in turn.

The question, as written, did not provide a timeline, or sequence of events associated with the

plant conditions; rather, it provided an applicant a list of current plant conditions as they applied

to the question, at a given point in time. The applicant is informed that a SGTR is in progress,

that operators are performing procedure 2-EOP-04, STEAM GENERATOR TUBE RUPTURE

SGTR, and that initial cooldown for S/G isolation is in progress using SBCS. Although we do

not know a specific time for the above conditions, the information provided means that the

operators performing the initial cooldown are somewhere between step 10 of EOP-04 and step

of EOP-04. The high-level action statements of these steps are as follows:

10. INITIATE lowering RCS Thot to less than 510 °F using SBCS.

11. DEPRESSURIZE the RCS in preparation for isolating the affected S/G:

2. WHEN permissive conditions are MET during a controlled cooldown,

THEN BLOCK automatic MSIS and SIAS actuation signals as follows:

13. VERIFY circulating water flow to the main condenser.

14. STABILIZE the Secondary Plant per Appendix X, Secondary Plant Post

Trip Actions, Section 2

15. IF a LOOP has occurred, THEN PERFORM the following.

16. DETERMINE the MOST affected S/G by evaluating the following

indications:

17. WHEN RCS Thot is less than 510 °F, THEN ISOLATE the MOST

affected S/G per Appendix R, Steam Generator Isolation.

Therefore, given the information provided in the stem, the applicant can place the operators as

performing the initial cooldown, somewhere between steps 10 and 17 of EOP-04. This is what

is meant by the term currently in the first part question statement.

The final bulleted information in the given statement of the question is that both AFAS-1 and

AFAS-2 interlocks have actuated. Note that no other information is given concerning any

additional operator actions taken to reset the AFAS signals or isolate the C AFW pump steam

supply from the ruptured S/G. Recall that Appendix E of NUREG-1021 stated: you should

assume that no operator actions have been taken, unless the stem of the question or the

answer choices specifically state otherwise. A correct application of this section of Appendix E

is that no other operator actions have been taken except those that would have been

directed in steps 10-17 of EOP-04. To assert otherwise would be to be making an assumption

unsupported by Appendix E requirements or the given information of question 99.

Therefore, the correct application of the given conditions of the question is that AFAS-1 and

AFAS-2 actuated at some point before where the operators are now in the procedure flowpath,

that is, somewhere between steps 10-17 of EOP-04. No other operator actions can be

assumed to have occurred. Because AFAS-1 and AFAS-2 actuation would have automatically

started the C AFW Pump, and because there are no operator actions in steps 10-17 to isolate

the C AFW Pump steam admission valves from the ruptured S/G, then the technically correct

determination is that a release is currently in progress as defined by procedure EPIP-02.

The contention above asserts that the operators would be using the operations hard card

procedure, 2-NOP-99, and completing various actions that were not specifically directed by the

EOP steps. This assertion cannot be considered correct for several reasons: (1) this is an

unsupported assumption as detailed in Appendix E quoted above; (2) the EOP steps never

direct the operators to take action in accordance with 2-NOP-99; and (3) moreover, EOP-04

step 17 specifically directs the operators to isolate the most affected S/G using Appendix R (not

NOP-99). In other words, the EOP requires the operators to complete the initial cooldown

before S/G isolation, and to conduct the isolation using Appendix R, not the operations hard

card (2-NOP-99).

The applicants assumptions are contrary to the given conditions of an AFAS actuation with no

other operator action. AFAS Actuation coupled with the other given plant conditions is

operationally valid in that given a large enough Steam Generator Tube Rupture (SGTR) that the

steam generators (SGs) would reach the AFAS setpoint and compete the 240 second timeout to

actuate prior to crew action. This was observed during simulator scenarios during this exam

that mirrored the conditions as given in this question. Although the NRC agrees with some of

the assertions in the above comment, and disagrees with others, it is sufficient to have shown

that they do not have a material bearing on the correct logical application of the NUREG-1021

requirements to this question.

To summarize again: it was a given condition in the question that AFAS-1 and AFAS2 had

actuated, this is an operationally valid occurrence during a SGTR, therefore 2C AFW Pump had

to be running since there were no cues in the question to indicate otherwise. With the 2C AFW

Pump running and the steam admission valve from the affected SG with a SGTR still open, this

is by the definition of EPIP-02, a Release to the environment and would be required to be

documented as such on the State Notification Form. A release was currently in progress.

Since a release cannot be simultaneously in progress and not in progress, there cannot be two

correct answers to the question. The NRC determined that the question, as written, was

technically accurate and the question had only one correct answer.

Post-Examination Comment #6: Job Performance Measure JPM A-2S and A-2R, Perform

RCS inventory Balance for the RO and SRO Candidates,

Comment

During administration of PSL L-20-1 NRC A-2S and A-2R; Perform RCS Inventory Balance for

the RO and SRO Candidates, an unintended typographical error was identified which resulted in

a math error in the JPM key. The station has corrected the typographical error in the calculation

in the JPM.

Station Recommendation: It is the stations request to update the two JPMs. [sic]

NRC Resolution

The licensees recommendation was accepted.

The typographical/arithmetic errors that were listed in the JPM key were corrected, and all

applicants were evaluated against the correct mathematical determination of the RCS leak rate.

Post-Examination Comment #7: Job Performance Measure JPM S-7, Start Containment

Purge/Respond to High Radiation- Unit 2

Comment

The JPM directs the candidate to perform a Containment Purge for Refueling Operations, in

accordance with 2-NOP-06.20, beginning with Section 4.2.1, Step 5. Performance Step 3 of the

JPM is identified as a critical step and states, ensure the Purge Mode selector switch is in the

Refuel position prior to fuel movement. This is contrary to 2-GOP-365, Refueling Sequencing

Guidelines. The Purge Mode selector switch position, in support of refueling, is maintained/

aligned per 2-GOP-365, Refueling Sequencing Guidelines, Section 4.0, step 31. Specifically,

step 31 states, ensure the following is complete prior to performing 0-NOP-67.05, Refueling

Operation. Ensure the Purge Mode Selector switch is selected to the Refuel position. During

refueling operations at PSL, this Purge Mode selector switch is positioned prior to refueling in

accordance with the procedure for Refueling Sequencing Guidelines.

Station Recommendation: It is the stations position that reference to the purge mode selector

switch in the task standard be removed from the JPM and performance step 3 be changed to

not critical.

NRC Resolution

The licensees recommendation was accepted.

Based on 2-GOP-365 procedural guidance that would ensure the Purge Mode Selector Switch

was in the proper position prior to fuel movement, the NRC concurs that step 2-NOP-06.20,

Section 4.2.1, Step 5 should not be a critical step for JPM S-7. The NRC agrees that

acceptable performance for JPM step 5 would either consist of positioning the Purge Mode

selector switch to REFUEL, or by verbalizing that the Purge Mode selector switch would be re-

positioned before commencing moving fuel in the Reactor. All applicants were evaluated

against this change to the JPM answer key.

SIMULATOR FIDELITY REPORT

Facility Licensee: St. Lucie Nuclear Plant

Facility Docket No.: 05000335, 05000389

Operating Test Administered: December 7 - 12, 2020

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.

No simulator fidelity or configuration issues were identified.

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