ML23242A052

07-Post Exam Comments

ML23242A052
Person / Time
Site:	South Texas
Issue date:	08/09/2023
From:	Heather Gepford NRC/RGN-IV/DORS/OB
To:	South Texas
References
Download: ML23242A052 (1)

Text

.......

Nuclear Operating Company -

Withhold from Public Disclosure Under 10 CFR 2.390 South Texas Project Electric Generating Station P.O. Box 289 Wadsworth. Texas 77483 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 1600 E. Lamar Boulevard Arlington, TX 76011-4511 South Texas Project Units 1 and 2 Docket Nos. STN 50-498, STN 50-499 Post-Examination Material August 21, 2023 NOC-AE-23003984 STI: 35504078 10 CFR 55 10 CFR 2.390 File No.: D43.02 This letter provides Post-Examination Materials for the NRC License Examim:1tion that was administered on July 17, 20*23, through August 9, 2023, at South Texas Project (STP), in accordance with NUREG-1021, Operator Licensing Examination _Standards for Power Reactors.

STP Nuclear Operating Company requests the NRC consider Attachment 2 to this letter as Exempt from Public Disclosure. None of the exam materials are to go to the Public Document Room (ADAMS) for at least two years.

There are no commitments in this letter.

The following Post-Examination material is being sent:

1. Written examination cover sheet with the original graded answer sheet and 1 clean copy of the answer sheet for each examinee. (Paper Copy)

2. Post-Exam Electronic Files Sent Via Box:

• LOT 27 NRC JPMs

• LOT 27 NRC Simulator Scenarios

• LOT 27 NRC Written Exam

• LOT 27 NRC Written Exam Analysis

• LOT 27 NRC Written Exam Applicant Comments After Exam

• LOT 27 NRC Written Exam Question Challenges by STP

• LOT 27 NRC Written Exam Questio_ns and Clarifications During Exam

• LOT 27 NRC Written Exam Seating Arrangement The completed Form 1.3-1, Examination Security Agreement, will be forwarded as soon as possible. -

Withhold from Public Disclosure Under 10 CFR 2.390 Withhold from Public Disclosure Under 10 CFR 2.390 NOC-AE-23003984 Page 2 of 2 If you have any questions, please contact Arthur Vest at (361) 972-8666 or me at (361) 972-4523.

Michael A. Fortner Manager, Operations Training Attachments:

1. Written examination cover sheets with the original graded answer sheet and 1 clean copy of the answer sheet for each examinee. (Paper Copy)

2. Post-Exam Electronic Files Sent Via Box cc:

John Kirkland Chief Examiner U. S. Nuclear Regulatory Commission 1600 E. Lamar Blvd.

Arlington, TX 76011-4511

• Attention: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 STP Records Management without Attachments 1 or 2 -

Withhold from Public Disclosure Under 10 CFR 2.390

LOT 27 NRC Exam Written Examination Analysis Bank #

Question #

Answer A

B C

D Total Missed

% Missed RO/SRO (11 Candidates) 3243 1

C 1

1 9

3311 2

C 1

1 9

3248 3

C 1

1 9

3312 4

D 1

3 4

36 3247 5

B 0

0 3318 6

B 5

5 45 1865 7

C 1

1 9

3242 8

B 0

0 3245 9

A 5

5 45 816 10 A

0 0

1021 11 B

1 1

9 3241 12 B

1 1

9 3240 13 C

0 0

2881 14 D

0 0

1742 15 C

1 1

9 3244 16 B

1 1

2 18 3099 17 D

1 1

9 3250 18 D

1 1

9 3254 19 A

5 5

45 3251 20 A

2 2

18 3255 21 A

1 1

9 3328 22 B

2 1

3 27 2817 23 C

1 1

2 18 2282 24 A

0 0

3253 25 A

1 1

9 3330 26 C

5 1

6 55 3271 27 C

1 1

9 3278 28 D

2 2

18 3293 29 D

5 5

45 3012 30 B

0 0

3275 31 B

3 1

4 36 3266 32 D

0 0

3264 33 D

1 1

9 3259 34 D

1 1

9 3279 35 C

2 2

18 3260 36 C

0 0

1801 37 A

0 0

3261 38 D

1 1

9 2957 39 B

0 0

2911 40 B

2 2

18 3262 41 A

1 1

9 3272 42 A

2 2

18 3319 43 B

3 3

27 3276 44 B

1 9

LOT 27 NRC Exam Written Examination Analysis Bank #

Question #

Answer A

B C

D Total Missed

% Missed 3256 45 D

2 2

18 3273 46 C

2 2

18 3011 47 D

1 1

9 3323 48 B

0 0

3274 49 D

1 1

9 3263 50 A

1 1

9 2618 51 A

0 0

3277 52 B

0 0

3265 53 B

1 1

9 2899 54 A

1 1

9 3286 55 C

1 1

9 3287 56 A

2 2

18 3324 57 D

1 2

3 27 3284 58 A

2 2

18 3325 59 D

4 4

36 3270 60 D

0 0

3285 61 C

1 1

9 3329 62 A

2 2

18 3268 63 D

1 1

9 2714 64 C

0 0

1940 65 C

0 0

2343 66 D

3 3

27 3258 67 D

0 0

2780 68 B

0 0

3257 69 C

1 1

9 3327 70 B

1 1

9 3281 71 C

0 0

3288 72 B

1 1

9 3291 73 C

2 2

18 3289 74 A

0 0

3290 75 D

1 1

1 9

SRO ONLY (6 Candidates) 3299 76 A

0 0

3331 77 D

3 3

50 3333 78 C

0 0

3326 79 B

1 1

17 3301 80 B

1 1

17 3303 81 A

0 0

3315 82 B

1 1

17 3307 83 A

1 1

2 33 3313 84 B

0 0

3321 85 D

1 1

17 3316 86 D

0 0

3306 87 D

2 2

33 3305 88 C

0 0

LOT 27 NRC Exam Written Examination Analysis Bank #

Question #

Answer A

B C

D Total Missed

% Missed 3308 89 C

0 0

3309 90 D

1 1

2 33 3317 91 A

0 0

3322 92 C

5 5

83 3100 93 B

0 0

2431 94 A

0 0

3332 95 C

2 2

33 3297 96 A

0 0

3298 97 C

2 2

33 2670 98 A

0 0

3295 99 B

0 0

3294 100 B

0 0

Question 26 6 Candidates missed this question for a 55% miss rate. 5 of 11 candidates selected A, 1 candidate selected D. The correct answer is C.

The question required the candidates to recall the maximum time allowed to reach MODE 3 for the primary to secondary leakage and the leakage change over time given in stem. 5 candidates did not determine the correct time frame. The question also asked the students to determine the amount of time it would take to exceed the tech spec limit based on conditions given in the stem. Only one candidate did not make the correct determination. The question is technically correct as written. The question and references are included in this document. No changes to the examination are necessary.

Question 77 3 of 6 SRO candidates missed this question for 50% miss rate. All candidates that missed this question selected B. The correct answer is D.

This question required the candidates to know actions contained in an Annunciator Response Procedure (ARP) and how the ARP would be implemented with the Emergency Operating Procedures (EOP). The candidates that missed the question did not know that the ARP specifically required completion of additional steps in the ARP in conjunction with the EOPs. The question is technically correct as written.

The question and references are included in this document. No changes to the examination are necessary.

LOT 27 NRC Exam Written Examination Analysis Question 92 5 of 6 candidates missed this question for an 83% miss rate. All candidates that missed this question selected D. The correct answer is C.

The 1st half of the question required the students to recall the maximum allowable overload setting for the Refueling Machine Hoist per TRM LCO 3.9.6. The 2nd half of the question required the students to recall the bases for the TRM associated with the Refueling Machine TRM LCO. All 6 candidates got the first portion of the question correct. 5 of 6 candidates did not know that the reactor vessel was a part of the bases for the overload. The question is technically correct as written. The question and references are included in this document. No changes to the examination are necessary.



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0POP04-RC-0004 Steam Generator Tube Leakage Rev. 36 Page 35 of 118 This Procedure is Applicable in Modes 1, 2, 3 and 4 Addendum 6 Recommended Response Times Addendum 6 Page 1 of 1 CAUTION Post Reactor Shutdown conditions in the primary to secondary leakage (RCS temperature and pressure decreasing and SG pressure increasing) may reduce the SG Tube Leakage Rate. Plant Shutdown and Cooldown Rates should be based on the initial or increased leakage and NOT reduced leakage due to the Post Shutdown conditions.

(1) Response times are the maximum times allowed. Power Reduction and Mode Change(s) may be completed in less time.

(2) Continued Operations per Plant Management direction. Refer to 0PGP03-ZO-0041, Action For Monitoring Primary to Secondary Leakage.

(3) Loss of Continuous Radiation Monitoring as defined in 0PGP03-ZO-0041, Action for Monitoring Primary to Secondary Leakage.

(4) With a continued increase in leakage rate over 30 minute time interval per the next column.

Action Level Leak Rate Increasing Leak Rate Response Times (1) 3

> 75 gpd (4)

Rate of inc > 30 gpd/hr Reduce Rx PWR to

< 50% in 1 hr AND Mode 3 in the next 2 hr 3

> 75 gpd AND Loss of Continuous Radiation Monitoring (3)

NA Reduce Rx PWR to

< 50% in 1 hr AND Mode 3 in the next 2 hr 3

> 150 gpd NA Mode 3 < 6 hr 2

> 75 gpd (4)

Rate of inc < 30 gpd/hr Mode 3 < 24 hr 1

< 75 gpd AND Loss of Continuous Radiation Monitoring (3)

N/A Continued Operations (2) 1

> 30 gpd N/A Continued Operations (2)

Increased Monitoring

> 5 gpd N/A Continued Operations (2)

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System operational leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 gpm UNIDENTIFIED LEAKAGE,

c. 150 gallons per day of primary-to-secondary leakage through any one steam generator,

d. 10 gpm IDENTIFIED LEAKAGE from the Reactor Coolant System, and

e. 0.5 gpm leakage per nominal inch of valve size up to a maximum of 5 gpm at a Reactor Coolant System pressure of 2235 r 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.*

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

D With any PRESSURE BOUNDARY LEAKAGE, isolate the affected component, pipe, or vessel from the RCS by use of a closeG manual valve, closed and de-activated automatic valve, blind flange, or check valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

E With Reactor Coolant System operational UNIDENTIFIED or IDENTIFIED LEAKAGE greater than the above limits, reduce leakage to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be inat least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

F With primary-to-secondary leakage not within the limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

G With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

• Test pressures less than 2235 psig but greater than 150 psig are allowed. Observed leakage shall be adjusted for the actual test pressure up to 2235 psig assuming the leakage to be directly proportional to pressure differential to the one-half power.

SOUTH TEXAS - UNITS 1 & 2 3/4 4-20 Unit 1 - Amendment No. 83, 90, 164, 225 Unit 2 - Amendment No. 77, 154, 210

6/28/2023 Print Date STP LOT-27 NRC SRO EXAM Unit 1 is at 100% power.

x CCP 1B trips on overcurrent.

x The crew is performing actions 0POP09-AN-04M8 Window F-3, CHG FLOW HI/LO.

x CCP 1A can NOT be started.

x RCP 1C Thermal Barrier cooling is unavailable.

x RCP 1C #1 Seal Inlet Temperature is 150ºF.

The Unit Supervisor will direct the crew to trip the reactor, verify turbine tripped, secure RCP 1C and ____(1)____.

The crew ____(2)____ start the PDP.

A. (1) GO TO 0POP05-EO-EO00, Reactor Trip or Safety Injection ONLY (2) should NOT B. (1) GO TO 0POP05-EO-EO00, Reactor Trip or Safety Injection ONLY (2) should C. (1) PERFORM 0POP05-EO-EO00, Reactor Trip or Safety Injection while continuing the Annunciator Response Procedure (2) should NOT D. (1) PERFORM 0POP05-EO-EO00, Reactor Trip or Safety Injection while continuing the Annunciator Response Procedure (2) should Answer: D (1) PERFORM 0POP05-EO-EO00, Reactor Trip or Safety Injection while continuing the Annunciator Response Procedure; (2) should Exam Bank No.: 3331 Last used on an NRC exam: Never SRO Sequence Number:77 Page 3 of 50

6/28/2023 Print Date STP LOT-27 NRC SRO EXAM K/A Catalog Number:

APE022 G2.4.16 Tier: 1 Group/Category: 1 Loss of Reactor Coolant Makeup: Knowledge of emergency and abnormal operating procedures implementation hierarchy and coordination with other support procedures or guidelines, such as operating procedures, abnormal operating procedures, or severe accident management guidelines.

STP Lesson: LOT 505.01 Objective Number: 92109 Given a plant conditon, DESCRIBE and/or INTERPRET the requirements and/or limits of a precaution or step of a referenced procedure.

Attached Reference

Reference:

0POP09-AN-04M8 pg. 74; 0POP01-ZA-0018, step 4.26.4, pg. 18

Attachment:

Source: New Distractor Justification A: INCORRECT. Plausible because this is the normal transition requirements but the ARP specifically calls out performing additional steps to start the PDP while performing EO00 and the RCP #1 seal inlet temperature is high but not high enough preclude starting the PDP.

B: INCORRECT. Plausible because this is the normal transition requirements but the ARP specifically calls out performing additional steps to start the PDP while performing EO00. Part (2) is correct.

C: INCORRECT. Plausible because the RCP #1 seal inlet temperature is high but not high enough preclude starting the PDP. Part (1) is correct.

D: CORRECT. Per 0POP09-AM-04M8 F-3 the crew should perform 0POP05-EO-EO00 actions but also continue performing the next step of the ARP which is to start the PDP. This is further supported by 0POP01-ZA-0018. The PDP is allowed be be started per the ARP provided all RCP

1. 1 Seal inlet temperatures are less than 230 degrees F. RCP 1C is the only RCP that does not have thermal barrier cooling and it is < 230 degrees per the question stem so the crew should start the PDP per the ARP.

Question Level: H Question Difficulty 4

Justification:

Student must analyze given plant conditions and determine proper actions.

SRO Importance: 4.4 NRC Reference Req'd

Attachment:

Modified From Exam Bank No.: 3331 10CFR Reference or SRO Objective:

55.43(b)(5)

SRO Justification:

Student must know the implementation heirachy of EOPs and abnormal procedures.

Page 4 of 50

0POP09-AN-04M8 Rev. 45 Page 74 of 83 ANNUNCIATOR LAMPBOX 04M8 RESPONSE INSTRUCTIONS CHG FLOW HI/LO Subsequent Actions:

(continued) d) IF CCW Thermal Barrier cooling AND RCP seal injection are lost, THEN PERFORM the following:

1. TRIP the Reactor

2. ENSURE Main Turbine tripped.

3. STOP affected RCP(s) within 1 minute.

4. PERFORM 0POP05-EO-EO00, Reactor Trip or Safety Injection WHILE CONTINUING with the next step.

e) IF all RCP # 1 Seal inlet temperatures are LESS THAN 230ºF, THEN PERFORM the following:

1. ENSURE RECIRC HCV-0285 is full open.

2. START PDP.

3. THROTTLE CLOSE RECIRV HCV-0285 to restore Seal flow to between 6 and 13 GPM to each RCP.

f) PERFORM 0POP04-RC-0002, Reactor Coolant Pump Off Normal, as resources permit.

g) PLACE Excess Letdown in service as directed by SM/US per 0POP02-CV-0004, Chemical and Volume Control System Subsystem.

3) CHECK for excessive RCS leakage by comparing charging and letdown flows.

Page 2 of 4 04M8-F-3 CHG FLOW HI/LO

0POP01-ZA-0018 Rev. 26 Page 18 of 47 Emergency Operating Procedure User's Guide CAUTION IF the AFW cross-connects are opened, THEN care SHALL be taken NOT to exceed 675 gpm flow on any AFW pump to prevent runout of the pump.

4.26.3 IF specific EOP SG(s) levels OR minimum AFW flow requirements are met, THEN the following actions can be taken:

x Reset the AFW system (including resetting the SI signal, SG LO-LO LVL and ESF Load Sequencers), open the AFW cross-connect valves, and decrease the number of running AFW pumps.

x Reset the AFW system (including resetting the SI signal and SG LO-LO LVL, if closure of the AFW OCIV(s) is needed) and manually control AFW flow.

x Isolate AFW flow to a faulted OR ruptured SG by placing the appropriate AFW pump in Pull-To-Lock.

4.26.4 Actions should be taken per Off Normal Operating Procedures and Annunciator Response Procedures that DO NOT conflict with the actions of the EOPs if adequate resources are available. The Off Normal Operating Procedure or Annunciator Response Procedure should be entered and procedure steps followed. (e.g., IF during the performance of the EOPs there are indications of abnormal RCP conditions, THEN the RCP Off Normal Operating Procedure SHOULD be entered.)

4.26.5 IF necessary to reset SI, Phase A, or ESF Load Sequencers to mitigate the consequences of the accident, prevent equipment damage OR protect the health and safety of plant personnel, THEN the SM/US should direct these actions prior to direction by the EOPs.

6/28/2023 Print Date STP LOT-27 NRC SRO EXAM Per the Technical Requirements Manual (TRM) the Refueling Machine must have an automatic overload cutoff set at a maximum of ____(1)____ pounds.

This provides protection for the ______(2)______ from excessive lifting force in the event they are inadvertently engaged during lift operations.

A. (1) 2650 (2) core internals and reactor vessel B. (1) 2650 (2) core internals only C. (1) 3250 (2) core internals and reactor vessel D. (1) 3250 (2) core internals only Answer: C (1) 3250; (2) core internals and reactor vessel Exam Bank No.: 3322 Last used on an NRC exam: Never SRO Sequence Number:92 Page 33 of 50

6/28/2023 Print Date STP LOT-27 NRC SRO EXAM K/A Catalog Number:

034 K4.01 Tier: 2 Group/Category: 2 Knowledge of Fuel Handling Equipment design features and/or interlocks that provide for the following: Fuel protection from binding and dropping.

STP Lesson: LOT 201.43 Objective Number: 33172 Given the topic or title of a specification included in the Technical Specifications or Technical Requirements Manual (TRM) DESCRIBE the general requirements of the specification to include the components or administrative requirements affected, limitations, major time frames involved, major surveillances in order to comply, and the basis for the specification/requirement.

Attached Reference

Reference:

TRM 3.9.6 and Bases; 0POP08-FH-0001

Attachment:

Source: New Distractor Justification A: INCORRECT. Plausible because the Refueling machine has a load select switch that provides variable overload setpoints that are more conservative than required by the TRM (2650 pounds is the nominal setpoint for position 3 of the switch). Part 2 is correct.

B: INCORRECT. Plausible because the Refueling machine has a load select switch that provides variable overload setpoints that are more conservative than required by the TRM (2650 pounds is the nominal setpoint for position 3 of the switch) and the core internals is part of the bases but not all of it.

C: CORRECT. TRM 3.96 requires an automatic overload cutoff of a maxixum of 3250 pounds and the Bases states that the overload protects both the core internals and the reactor vessel from excessive lifting forces in the event they are inadvertently engaged during lift operations.

D: INCORRECT. Plausible because the core internals is part of the bases but not all of it. Part 1 is correct.

Question Level: F Question Difficulty 3

Justification:

Student must recall the content of the TRM and Bases.

SRO Importance: 3.3 NRC Reference Req'd

Attachment:

Modified From Exam Bank No.: 3322 10CFR Reference or SRO Objective:

55.43(b)(7)

SRO Justification:

This is unique to SRO Position.

Page 34 of 50


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