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| issue date = 10/04/2013
| issue date = 10/04/2013
| title = STP 2013-09 Post Examination Comments
| title = STP 2013-09 Post Examination Comments
| author name = Osterholtz C C
| author name = Osterholtz C
| author affiliation = NRC/RGN-IV/DRS/OB
| author affiliation = NRC/RGN-IV/DRS/OB
| addressee name =  
| addressee name =  
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{{#Wiki_filter:POST-EXAM APPLICANT COMMENTS STP LOT 19 NRC WRITTEN EXAM 9/26/2013 Question #28:
{{#Wiki_filter: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)
(missed by 10 applicants, 8 chose 'B', 2 chose 'C')
 
Given the following:
Given the following:
A Loss of Offsite Power (LOOP) has occurred on Unit 1 while operating at 100% power.
* 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 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:
* The Secondary RO also notes the following AFW Pump flows:
AFW Pump 11
* AFW Pump 11           ..570 gpm
--..570 gpm AFW Pump 12
* AFW Pump 12           ..600 gpm
--..600 gpm AFW Pump 13
* AFW Pump 13           ..0 gpm
---..0 gpm TDAFW Pump
* TDAFW Pump             ..560 gpm Which of the following describes the reason for the given indications?
--..560 gpm
Steam Generator C.
 
A.     OCIV, MOV-0065, failed to receive an actuation signal.
Which of the following describes the reason for the given indications?
B.     AFW REG, FV-7523, failed to receive a control signal from QDPS.
Steam Generator 'C'-.
C.     NARROW RANGE level has not yet lowered to less than 20%.
A. OCIV, MOV-0065, failed to receive an actuation signal.
D.     AFW AUTO FLOW CONT RESET pushbutton has NOT been depressed.
B. AFW REG, FV
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:
-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).


POST-EXAM APPLICANT COMMENTS STP LOT 19 NRC WRITTEN EXAM 9/26/2013 Valve Control High Limit (640 GPM)
Valve Control Low Limit (550 GPM)
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)
(ES-403, D.1.b, 3rd bullet)
NRC RESOLUTION: Recommendation accepted, both A and B to be accepted as correct.


NRC RESOLUTION:
POST-EXAM APPLICANT COMMENTS STP LOT 19 NRC WRITTEN EXAM 9/26/2013
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 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 Injection.
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.
The crew is performing 0POP05-EO-EO00, Reactor Trip or Safety Injection.
Based on the following conditions of the Steam Generators and Containment; Steam                    A                  B                C                  D Generators Pressure        1095 psig            1085 psig        1090 psig          1010 psig Slowly Lowering      Slowly Lowering  Slowly Lowering    Slowly Lowering Level            20% NR              19% NR            29% NR              31% NR Slowly Rising        Slowly Rising    Stable              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)


Based on the following condition s of the Steam Generators and Containment
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.
;  Steam Generators A B C D Pressure 10 95 psig Slowly Lowering 10 85 psig Slowly Lowering 10 90 psig Slowly Lowering 10 10 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:
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)
(missed by 4 applicants, 2 chose 'A' and 2 chose 'C')
Unit 1 is operating at 100% Power.
Unit 1 is operating at 100%
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.
Power. The crew has implemented 0POP04
Steam                     A                   B                 C                   D Generators Steam Line         1.8E-2 uCi/cc       1.5E-2 uCi/cc     1.4E-2 uCi/cc       3.9E-1 uCi/cc Radiation Blowdown           3.1E-4 uCi/cc       2.4E-4 uCi/cc     2.3E-4 uCi/cc       4.6E-2 uCi/cc Radiation N-16                   9.0 gpd             0.2 gpd           0.1 gpd             77.0 gpd Monitors
-RC-0004, Steam Generator Tube Leakage, due to the following current Radiation Monitor Readings given to the Unit Supervisor for the Steam Generators.
* Chemistry reports total current primary to secondary leak rate is 75 gpd.
Steam Generators A B C D Steam Line Radiation 1.8 E-2 uCi/cc 1.5 E-2 uCi/cc 1.4 E-2 uCi/cc 3.9E-1 uCi/cc Blowdown Radiation 3.1 E-4 uCi/cc 2.4 E-4 uCi/cc 2.3 E-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
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.
A. Only Steam Generator 'D' has a tube leak  
B. Only Steam Generator D has a tube leak - Increased radiation dose to plant workers AND radiological release to the environment.
- Increased radiation dose to plant workers ONLY. B. Only Steam Generator 'D' has a tube leak  
C. Steam Generators A and D have tube leaks - Increased radiation dose to plant workers ONLY.
- Increased radiation dose to plant workers AND radiological release to the environment.
D. Steam Generators A and D have tube leaks - Increased radiation dose to plant workers AND radiological release to the environment.
C. Steam Generators 'A' and 'D' have tube leaks  
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.
- Increased radiation dose to plant workers ONLY. D. Steam Generators 'A' and 'D' have tube leaks  
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.
- Increased radiation dose to plant workers AND radiological release to the environment.
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.
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)
(ES-403, D.1.b, 1st bullet)


POST-EXAM APPLICANT COMMENTS STP LOT 19 NRC WRITTEN EXAM 9/26/2013 NRC RESOLUTION:
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.
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
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)
-ZO-0012, Operations Troubleshooting Process, which of the following is correct concerning the troubleshooting process?
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).
: 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.
: 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
: 3. Operations Manager approval is required to enter into 0POP01-ZO-0012, Operations Troubleshooting Process.
-ZO-0012, Operations Troubleshooting Process. 4. Shift Manager approval is required to enter into 0POP01
: 4. Shift Manager approval is required to enter into 0POP01-ZO-0012, Operations Troubleshooting Process.
-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.
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.
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)
(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.


NRC RESOLUTION: Recommendation denied.
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
Procedure 0POP01-ZO-0012 contains a specific provision for allowing troubleshooting on operable equipment as specified in the


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


LOT 19 NRC EXAM ANALYSIS  
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).
- 9/2 6/2013   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
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.
- 9/2 6/2013  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 6 7  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
- 9/2 6/2013  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 2 177  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 - 1 3 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.
No changes to the exam are warranted.
RO #22 - 3 candidates chose distracter "A"
RO #22 - 3 candidates chose distracter A; 7 candidates chose distracter C; 1 candidate chose distracter D; (correct answer is B).
; 7 candidates chose distracter "
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.
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.
No changes to the exam are warranted.


LOT 19 NRC EXAM ANALYSIS  
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).
- 9/2 6/2013 RO #48 - 8 candidates chose distracter "A"
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.
; 2 candidates chose distracter "
No changes to the exam are warranted.
B"; 1 candidate chose distracter "
RO #74 - 3 candidates chose distracter A; 7 candidates chose distracter B (correct answer is D).
D"; (correct answer is "
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.
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.
. No changes to the exam are warranted.
SRO #90 - 7 candidates chose distracter B; 1 candidate chose distracter D (correct answer is C).
RO #74 - 3 candidates chose distracter "A"
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.
; 7 candidates chose distracter "
Recommend changing answer key to allow both B & C to be the correct answers.
B" (correct answer is "
SRO #95 - 6 candidates chose distracter A; 2 candidates chose distracter C (correct answer is D).
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
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.
S RO #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.}}
No changes to the exam are warranted.}}

Latest revision as of 11:31, 4 November 2019

STP 2013-09 Post Examination Comments
ML13310B640
Person / Time
Site: South Texas  STP Nuclear Operating Company icon.png
Issue date: 10/04/2013
From: Clyde Osterholtz
Operations Branch IV
To:
South Texas
laura hurley
References
Download: ML13310B640 (12)
v • d • e


Text

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 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 Valve Control High Limit (640 GPM)

Valve Control Low Limit (550 GPM)

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.

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 Injection.

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

Based on the following conditions of the Steam Generators and Containment; Steam A B C D Generators Pressure 1095 psig 1085 psig 1090 psig 1010 psig Slowly Lowering Slowly Lowering Slowly Lowering Slowly Lowering Level 20% NR 19% NR 29% NR 31% NR Slowly Rising Slowly Rising Stable 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 A B C D Generators Steam Line 1.8E-2 uCi/cc 1.5E-2 uCi/cc 1.4E-2 uCi/cc 3.9E-1 uCi/cc Radiation Blowdown 3.1E-4 uCi/cc 2.4E-4 uCi/cc 2.3E-4 uCi/cc 4.6E-2 uCi/cc Radiation N-16 9.0 gpd 0.2 gpd 0.1 gpd 77.0 gpd Monitors

  • 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.

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