ML13310B640

STP 2013-09 Post Examination Comments

ML13310B640
Person / Time
Site:	South Texas
Issue date:	10/04/2013
From:	Osterholtz C C Operations Branch IV
To:	South Texas
laura hurley
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Download: ML13310B640 (12)

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POST-EXAM APPLICANT COMMENTS STP LOT 19 NRC WRITTEN EXAM 9/26/2013 Question #28:

(missed by 10 applicants, 8 chose 'B', 2 chose 'C')

Given the following:

A Loss of Offsite Power (LOOP) has occurred on Unit 1 while operating at 100% power.

The secondary RO reports that all Motor Driven Auxiliary Feedwater (AFW) pumps and the Turbine Drive Auxiliary Feedwater (TDAFW) Pump are running.

The Secondary RO also notes the following AFW Pump flows:

AFW Pump 11

--..570 gpm AFW Pump 12

--..600 gpm AFW Pump 13

---..0 gpm TDAFW Pump

--..560 gpm

Which of the following describes the reason for the given indications?

Steam Generator 'C'-.

A. OCIV, MOV

-0065, failed to receive an actuation signal.

B. AFW REG, FV

-7523, failed to receive a control signal from QDPS.

C. NARROW RANGE level has not yet lowered to less than 20%.

D. AFW AUTO FLOW CONT RESET pushbutton has NOT been depressed.

Keyed answer: A

COMMENT: Both the Reg Valve control signal and the control board flow indication originate in QDPS Auxiliary Processing Cabinet (APC) #2. If QDPS APC #2 was not sending an output to the Reg Valve due to an internal problem, then it would be reasonable to expect the same processing cabinet was not sending an output to the control board indication and it would be reading '0' gpm. See figure below from lesson LOT202.44 handout

POST-EXAM APPLICANT COMMENTS STP LOT 19 NRC WRITTEN EXAM 9/26/2013 RECOMMENDATION:

Accept both 'A' and 'B' as correct.

The "B" distracter does not specify why the valve did not receive a control signal; therefore it is possible for different assumptions to be made.

QDPS Engineer Mike Crutcher confirmed that if the QDPS APC #2 was out of service, then the flow indication would read "0" since the output to the indicator would be 0 volts (see excerpt below from the QDPS vendor manual).

(ES-403, D.1.b, 3rd bullet)

NRC RESOLUTION:

Recommendation accepted, both 'A' and 'B' to be accepted as correct.

Valve Control High Limit (640 GPM)Valve Control Low Limit (550 GPM)

POST-EXAM APPLICANT COMMENTS STP LOT 19 NRC WRITTEN EXAM 9/26/2013

POST-EXAM APPLICANT COMMENTS STP LOT 19 NRC WRITTEN EXAM 9/26/2013 Question #79 (missed by 6 applicants, all chose 'A')

Unit 1 was operating at 100% power when an event occurred that tripped the reactor and initiated a Safety Injecti on.

The crew is performing 0POP05-EO-EO00, Reactor Trip or Safety Injection.

Based on the following condition s of the Steam Generators and Containment

Steam Generators A B C D Pressure 1095 psig Slowly Lowering 1085 psig Slowly Lowering 1090 psig Slowly Lowering 1010 psig Slowly Lowering Level 20% NR Slowly Rising 19% NR Slowly Rising 29% NR Stable 31% NR Slowly Lowering AFW Flow 150 gpm 150 gpm 50 gpm 50 gpm Containment Pressure 3.2 psig - Rising Temperature 130ºF - Rising Humidity 110ºF-dew point - Rising Which of the following procedures should the Unit Supervisor perform next?

A. 0POP05-EO-EO20, Faulted Steam Generator Isolation B. 0POP05-EO-EO30, Steam Generator Tube Rupture C. 0POP05-EO-EO10, Loss of Reactor or Secondary Coolant D. 0POP05-EO-FRZ1, Response to High Containment Pressure Keyed answer: C COMMENT: All steam generator pressures are obviously lower than normal post

-trip pressure (~1185 psig) which could be indicative of a small steam break without any data for containment radiation.

RECOMMENDATION: Accept 'A' or 'C' as correct.

If a small steam break is occurring in containment, steam pressures and containment conditions would indicate as given and E20 entry would be required and cannot be ruled out without being given containment radiation

. If a SBLOCA is in progress that is large enough to cause ECCS injection flow resulting in an RCS cooldown, thus lowering SG pressures, which would also give the indicated trends in steam generator pressures and containment conditions and E10 entry would be required

. (ES-403, D.1.b, 1st bullet)

POST-EXAM APPLICANT COMMENTS STP LOT 19 NRC WRITTEN EXAM 9/26/2013 NRC RESOLUTION:

Change correct answer to 'A'. There is no indication of an RCS leak given no escalated radiation levels inside containment. Therefore

, the stem information only provides justification for a faulted steam generator.

POST-EXAM APPLICANT COMMENTS STP LOT 19 NRC WRITTEN EXAM 9/26/2013 Question #84:

(missed by 4 applicants, 2 chose 'A' and 2 chose 'C')

Unit 1 is operating at 100%

Power. The crew has implemented 0POP04

-RC-0004, Steam Generator Tube Leakage, due to the following current Radiation Monitor Readings given to the Unit Supervisor for the Steam Generators.

Steam Generators A B C D Steam Line Radiation 1.8E-2 uCi/cc 1.5E-2 uCi/cc 1.4E-2 uCi/cc 3.9E-1 uCi/cc Blowdown Radiation 3.1E-4 uCi/cc 2.4E-4 uCi/cc 2.3E-4 uCi/cc 4.6E-2 uCi/cc N-16 Monitors 9.0 gpd 0.2 gpd 0.1 gpd 77.0 gpd Chemistry reports total current primary to secondary leak rate is 75 gpd. Which Steam Generator(s) have tube leaks and what are the radiological hazards associated with this event

?

A. Only Steam Generator 'D' has a tube leak

- Increased radiation dose to plant workers ONLY. B. Only Steam Generator 'D' has a tube leak

- Increased radiation dose to plant workers AND radiological release to the environment.

C. Steam Generators 'A' and 'D' have tube leaks

- Increased radiation dose to plant workers ONLY. D. Steam Generators 'A' and 'D' have tube leaks

- Increased radiation dose to plant workers AND radiological release to the environment.

Keyed answer: B COMMENT: Per 0ERP01

-ZV-IN01, the definition for a radiological release is "Any radiological release from the plant that exceeds the EAL limits established for an Unusual Event." Since the given Steam Line radiation reading is only 10 times higher than normal, it would not be reasonable to assume the UE threshold has been exceeded

. RECOMMENDATION:

Accept 'A' and 'B' as correct. Since there is no context given for "radiological release", the SRO applicant could answer this question from an Emergency Director (EAL) standpoint or from a more generic radiological control standpoint.

(ES-403, D.1.b, 1st bullet)

POST-EXAM APPLICANT COMMENTS STP LOT 19 NRC WRITTEN EXAM 9/26/2013 NRC RESOLUTION:

Recommendation denied.

The licensee has one proceduralized definition for a radiological release as defined in their emergency plan. A senior reactor operator should be familiar with this definition and be able to apply it.

Answer B remains the only correct answer.

POST-EXAM APPLICANT COMMENTS STP LOT 19 NRC WRITTEN EXAM 9/26/2013

Question #90:

(missed by 8 applicants, 7 chose 'B' and 1 chose 'D')

Due to an emergent equipment condition, Plant Operations needs to generate a troubleshooting plan.

In accordance with 0POP01

-ZO-0012, Operations Troubleshooting Process, which of the following is correct concerning the troubleshooting process?

1. Component operation should ONLY be directed by approved plant procedures or Condition Report Operation Evaluations (CROE).

2. Troubleshooting should ONLY be performed on equipment that is already removed from service and/or inoperable.

3. Operations Manager approval is required to enter into 0POP01

-ZO-0012, Operations Troubleshooting Process. 4. Shift Manager approval is required to enter into 0POP01

-ZO-0012, Operations Troubleshooting Process. A. 1, 3 B. 2, 4 C. 1, 4 D. 2, 3 Keyed answer: C COMMENT: In accordance with the Plant Procedure Writers Guide, "should" is a recommended action vice "shall" which is a required action. Given that, item #2 in the stem of the question is also correct since it does not forbid performing troubleshooting on operable equipment.

RECOMMENDATION:

Accept 'B' or 'C' as correct. Procedure section 5.1 states "Troubleshooting should only be performed on operable systems which will not be rendered inoperable as determined by the Shift Supervisor, or on systems which have already been declared inoperable" Although worded slightly different from Item #2, they both functionally mean the same thing. If item #2 in the stem of the question had read "Troubleshooting shall ONLY be performed on equipment that is already removed from service and/or inoperable", the statement would be truly incorrect

.

(ES-403, D.1.b, 1st bullet)

NRC RESOLUTION: Recommendation denied.

Procedure 0POP01-ZO-0012 contains a specific provision for allowing troubleshooting on operable equipment as specified in the

"recommendation

" section above. This renders choice 2 in the stem as clearly incorrect. Answer C remains the only correct answer.

LOT 19 NRC EXAM ANALYSIS

- 9/26/2013 Bank # RO# SRO# ANSWER A B C D Total Missed 2225 1 C 0 2207 2 B 0 2208 3 B 0 19 4 A 1 1 52 5 D 0 2219 6 A 1 1 381 7 C 0 2221 8 D 0 2160 9 D 4 4 1296 10 B 0 1577 11 B 0 1651 12 D 0 2222 13 C 13 13 1820 14 C 0 1863 15 D 0 2224 16 D 1 1 2150 17 A 1 1 2228 18 A 1 1 2 2152 19 C 4 1 5 2153 20 D 4 4 2156 21 D 1 1 2157 22 B 3 7 1 11 2165 23 C 3 3 2168 24 A 1 1 2170 25 D 1 1 2171 26 D 7 1 8 2173 27 B 0 2232 28 A/B 2 2 2210 29 D 5 1 6 2213 30 C 1 3 4 164 31 B 0 278 32 C 1 1 854 33 B 1 4 5 32 34 B 0 82 35 D 1 1 2 432 36 C 3 3 478 37 B 0 490 38 B 1 2 3 492 39 C 1 1 2 2220 40 B 2 5 7 922 41 C 0 1077 42 A 0 1330 43 A 1 1 LOT 19 NRC EXAM ANALYSIS

- 9/26/2013 Bank # RO# SRO# ANSWER A B C D Total Missed 1401 44 B 0 1559 45 A 1 2 3 1665 46 D 2 1 3 389 47 B 0 1700 48 C 8 2 1 11 2231 49 B 1 1 2167 50 A 0 2175 51 A 0 2176 52 D 2 1 3 2179 53 A 0 2186 54 A 4 4 2189 55 C 1 1 2190 56 B 1 1 2 96 57 D 0 2192 58 A 0 2193 59 B 3 3 2194 60 C 0 2195 61 C 1 3 4 2196 62 D 6 6 2197 63 A 1 1 2198 64 A 1 1 3 5 2199 65 B 2 2 2200 66 B 2 7 9 2201 67 B 3 1 4 2202 68 D 2 1 3 2203 69 A 0 2204 70 D 0 2205 71 C 4 4 2209 72 A 0 2206 73 C 2 1 3 2071 74 D 3 7 10 2227 75 B 3 6 9 2181 76 D 1 1 2 2178 77 D 3 3 1627 78 D 0 2154 79 A 5 5 2155 80 B 1 2 3 2158 81 C 0 2159 82 B 0 2218 83 A 0 2161 84 B 2 2 4 2162 85 D 1 1 2216 86 B 1 1 2182 87 D 3 1 4 LOT 19 NRC EXAM ANALYSIS

- 9/26/2013 Bank # RO# SRO# ANSWER A B C D Total Missed 2211 88 C 1 1 2180 89 B 1 1 2223 90 C 7 1 8 2163 91 A 1 2 3 2164 92 B 0 2174 93 B 3 3 2177 94 A 0 2226 95 D 6 2 8 2229 96 D 5 5 2187 97 B 5 5 2188 98 B 1 1 1604 99 C 0 1732 100 C 1 1 RO #13 - 13 candidates chose distracter "A" (correct answer is "

C"). Question is based on Reactor Trip set point for RCP Under Frequency and the effects this has on Reactor Trip Breakers as well as RCP Breakers. The set point is 57.2 HZ. The incorrect distracter chosen indicates a lack of knowledge of the Reactor Trip set point of 57.2 HZ and NOT how the Reactor Trip Breakers and RCP Breakers respond

. Training is given on Reactor Trip setpoints in LOT 201.20, SSPS.

No changes to the exam are warranted.

RO #22 - 3 candidates chose distracter "A"

7 candidates chose distracter "

C"; 1 candidate chose distracter "

D"; (correct answer is "B"). Question is based on Westinghouse owners Group Background Documents. The NOTE referred to in the question is used in several Emergency Procedures when a cooldown is performed without RCPs running.

All distracters chosen indicate a lack of knowledge for the basis as to why it is not good to have a void in the Reactor Vessel Upper Head region during a depressurization without RCPS running. Training is given on Emergency Procedures and their basis during LOT 504, Emergency Procedure Training.

No changes to the exam are warranted.

LOT 19 NRC EXAM ANALYSIS

- 9/26/2013 RO #48 - 8 candidates chose distracter "A"

2 candidates chose distracter "

B"; 1 candidate chose distracter "

D"; (correct answer is "

C"). Question is based on what causes DRPI and Rod Control Urgent Alarms and how it would affect the ability of the operator to monitor Control Rod Position. Distracter "A" was chosen by most of the applicants and indicates a lack of knowledge on what would actually affect the ability of the operator to monitor Control Rod Position. Training is given on DRPI in LOT 201.19 and Rod Control in LOT 201.18

. No changes to the exam are warranted.

RO #74 - 3 candidates chose distracter "A"

7 candidates chose distracter "

B" (correct answer is "

D"). Question is based on the radiation hazards associated with new fuel assemblies. Applicants choosing distracters "A" & "B" indicate a lack of knowledge on how ionizing radiation affects the skin of the body. Training is given during fundamentals training in LOT 103

. No changes to the exam are warranted.

SRO #90 - 7 candidates chose distracter "

B"; 1 candidate chose distracter "D" (correct answer is "

C"). Question is based on knowledge of Troubleshooting process. After reviewing this question it was determined that terminology of "SHOULD' & "SHALL" was not used correctly being that this question came directly from the Troubleshooting procedure. See the Post Exam Student Comments.

Recommend changing answer key to allow both "

B" & "C" to be the correct answers.

SRO #95 - 6 candidates chose distracter "

A"; 2 candidates chose distracter "C" (correct answer is "

D"). Question is based on knowledge of how Pressurizer Pressure and Tavg can affect the OTDT channels. Most applicants picked distracter "B" which would indicate that the question was not completely analyzed. Training is given on Reactor Trip conditions in LOT 201.20, SSPS and on Technical Specifications in LOT 503.

No changes to the exam are warranted.