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| number = ML15259A292
| number = ML15259A292
| issue date = 09/16/2015
| issue date = 09/16/2015
| title = Watts Bar 2015-301 Draft SRO Written Exam
| title = 301 Draft SRO Written Exam
| author name =  
| author name =  
| author affiliation = NRC/RGN-II/DRS
| author affiliation = NRC/RGN-II/DRS
Line 15: Line 15:


=Text=
=Text=
{{#Wiki_filter:NRC Exam Legend Times are HH:MM:SS [ 00:00:00 - 23:59:59 ] ALARMS WINDOWS: LIT DARK LIGHT INDICATIONS: COLOR LIT DARK RED   GREEN   WHITE    
{{#Wiki_filter:NRC Exam Legend Times are HH:MM:SS [ 00:00:00 - 23:59:59 ]
: 76. Given the following conditions: - Unit 1 is at 28% power. - RCP SEAL LEAK OFF FLOW HI (100-D) is LIT. FR-62-24, SEAL LEAKOFF - HI RANGE - GPM reads 5.2 gpm. Subsequently, the following is observed on 1-M-5:  In accordance with BOTH 1-AOI-24 and TI-12.04, which ONE of the following completes the statement below?
ALARMS WINDOWS:
LIT                           DARK LIGHT INDICATIONS:
COLOR                 LIT                     DARK RED GREEN WHITE


76.
Given the following conditions:
    -  Unit 1 is at 28% power.
    -  RCP SEAL LEAK OFF FLOW HI (100-D) is LIT.
    -  1-FR-62-24, SEAL LEAKOFF - HI RANGE - GPM reads 5.2 gpm.
Subsequently, the following is observed on 1-M-5:
In accordance with BOTH 1-AOI-24 and TI-12.04, which ONE of the following completes the statement below?
The Unit Supervisor will direct ____(1)____.
The Unit Supervisor will direct ____(1)____.
NOTE: 1-AOI-24, RCP Malfunctions During Pump Operation TI-12.04, User's Guide for Abnormal and Emergency Operating Instructions ES-0.1, Reactor Trip Response A. a Unit 1 shutdown in accordance with 1-GO-5, Unit Shutdown From 30% Reactor Power to Hot Standby B. the OAC to trip the Unit 1 reactor and the CRO to perform further actions of 1-AOI-24 after the crew transitions to ES-0.1 C. the OAC to trip the Unit 1 reactor and the CRO to perform further actions of 1-AOI-24 while the OAC is performing his immediate actions D. the crew to maintain Unit 1 at power as a Unit Shutdown is NOT yet required and an operator to REFER TO Attachment 2, Immediate Shutdown Criteria CORRECT ANSWER:B DISTRACTOR ANALYSIS:   A. Incorrect: This distractor is incorrect and is tantamount in plausibility as the "D" distractor. If the US entered section 3.3 of 1-AOI-24 and then believed that seal leakoff flow were in excess of the criteria given for an initial controlled shutdown, then he would select an appropriate procedure (again because the unit was at a power of less than 30%, 1-GO-5 would be appropriate) and bring the Unit to mode 3. Therefore, the foregoing defends the plausibility of this distractor as this would be a correct procedural flowpath if the applicant did not understand that a trip setpoint were exceeded and that the seal leakoff criteria were exceeded. B. Correct: As mentioned, this distractor lists the correct action for the control room staff to take; this is, tripping the reactor when an immediate shutdown criterion is met and continuing with the steps of 1-AOI-24 after the transition to ES-0.1 is made. C. Incorrect: It is correct that a reactor trip would be required as an immediate shutdown criterion had been exceeded. Even if the Unit Supervisor did not initially recognize that such criteria had been exceeded and thus went to section 3.3; he would reach a step in such section which details "MONITOR RCP immediate shutdown required."  At this point he would enter section 3.2 and direct a reactor trip. Section 2.8, "Use of AOIs While in EOIs" of TI-12.04 contains: "3. When an AOI in effect directs a Reactor Trip and then the performance of the AOI should continue immediately following transition to ES-0.1."  Therefore, it would not be correct for the CRO to perform the steps of 1-AOI-24 in parallel with the immediate actions of the OAC (because the transition to ES-0.1 had not yet been made). It is plausible to believe this because if one did not recognize the restriction imposed by TI-12.04, one would logically interpret step 3. of 1-AOI-24 as directing exactly this. Step 3 of 1-AOI-24 reads: "TRIP the reactor, and GO TO E-0 Reactor Trip or Safety Injection, WHILE continuing with this instruction."  D. Incorrect: As seen in ARI-95-101, "Reactor Coolant Pumps," the setpoint for annunciator window 100-D is 4.8 gpm. Plant Operation has shown that leakoff values of approximately 2.3 gpm (this value is slightly variable) are normal. The question gives the fact that leakoff for the #1 seal for the #1 RCP is 5.2 gpm. Using Attachment 1 of 1-AOI-24, one may observe that the normal operating range for the #1 seal leakoff is between 1 to 5 gpm when the Unit is at normal operating pressure of 2235 psig. Therefore, it is fact that #1 seal leakoff is "high."  The stem of the question also gives the fact that 1-TI-62-3; "RCP 1 LWR BRG TEMP" is at 230°F. This is in excess of immediate trip criteria of 225°F (contained in Attachment 2 of 1-AOI-24). Two correct procedural avenues could be used to address this issue. Both will result in the same outcome. Firstly, the Unit Supervisor could immediately identify that RCP immediate trip criteria are met through memorization of the criteria of Attachment 2, "RCP Immediate Shutdown Criteria" and thus upon entry into 1-AOI-24 would select subsection 3.2, "RCP Tripped or Shutdown Required."  If the Unit Supervisor did not immediately identify that a trip of the RCP were required he would select section 3.3, "#1 Seal Leakoff Flow High."  Section 3.3 contains a decision step which selects whether a controlled shutdown (given that the unit is at 28% power a shutdown conducted per 1-GO-5 would be appropriate) is initially appropriate. If #1 seal leakoff is greater than or equal to 6.0 gpm then an initial shutdown is performed. If the seal leakoff is not in excess of this value, then the Unit Supervisor would assign an operator to "REFER TO Attachment 2" and thus utilize the Attachment to monitor for further degradation of the RCP seal package. If the Unit Supervisor did not recognize that an immediate trip of the RCP was required and because he had already bypassed the opportunity for plant shutdown (because seal leakoff were less than 6.0 gpm) then he would continue performing section 3.3 and thus maintain the Unit at power. The foregoing defends the plausibility of this distractor as this would be a correct procedural flowpath if the applicant did not understand that a trip setpoint were exceeded.
NOTE:     1-AOI-24, RCP Malfunctions During Pump Operation TI-12.04, Users Guide for Abnormal and Emergency Operating Instructions ES-0.1, Reactor Trip Response A. a Unit 1 shutdown in accordance with 1-GO-5, Unit Shutdown From 30%
Question Number: 76 Tier: 1 Group:   1 K/A: 015/017 Reactor Coolant Pump (RCP) Malfunctions 2.2 Equipment Control 2.2.44 Ability to interpret control room indications to verify the status and operation ofl a system, and understand how operator actions and directives affect plant and system conditions.
Reactor Power to Hot Standby B. the OAC to trip the Unit 1 reactor and the CRO to perform further actions of 1-AOI-24 after the crew transitions to ES-0.1 C. the OAC to trip the Unit 1 reactor and the CRO to perform further actions of 1-AOI-24 while the OAC is performing his immediate actions D. the crew to maintain Unit 1 at power as a Unit Shutdown is NOT yet required and an operator to REFER TO Attachment 2, Immediate Shutdown Criteria
Importance Rating: 4.2 4.4 10 CFR Part 55: (CFR: 41.5 / 43.5 / 45.12) 10CFR55.43.b: 10 CFR 55.43(b)(5) K/A Match: K/A is matched because the applicant is required to interpret the #1 RCP parameters to verify that the RCP is operating improperly and thus that the correct operator directives will cause the plant to be tripped and the RCP to be secured in accordance with the guidance of 1-AOI-24.
 
CORRECT ANSWER:                                                                 B DISTRACTOR ANALYSIS:
A. Incorrect: This distractor is incorrect and is tantamount in plausibility as the D distractor. If the US entered section 3.3 of 1-AOI-24 and then believed that seal leakoff flow were in excess of the criteria given for an initial controlled shutdown, then he would select an appropriate procedure (again because the unit was at a power of less than 30%,
1-GO-5 would be appropriate) and bring the Unit to mode 3. Therefore, the foregoing defends the plausibility of this distractor as this would be a correct procedural flowpath if the applicant did not understand that a trip setpoint were exceeded and that the seal leakoff criteria were exceeded.
B. Correct: As mentioned, this distractor lists the correct action for the control room staff to take; this is, tripping the reactor when an immediate shutdown criterion is met and continuing with the steps of 1-AOI-24 after the transition to ES-0.1 is made.
C. Incorrect: It is correct that a reactor trip would be required as an immediate shutdown criterion had been exceeded. Even if the Unit Supervisor did not initially recognize that such criteria had been exceeded and thus went to section 3.3; he would reach a step in such section which details MONITOR RCP immediate shutdown required. At this point he would enter section 3.2 and direct a reactor trip. Section 2.8, Use of AOIs While in EOIs of TI-12.04 contains: 3. When an AOI in effect directs a Reactor Trip and then the performance of the AOI should continue immediately following transition to ES-0.1. Therefore, it would not be correct for the CRO to perform the steps of 1-AOI-24 in parallel with the immediate actions of the OAC (because the transition to ES-0.1 had not yet been made). It is plausible to believe this because if one did not recognize the restriction imposed by TI-12.04, one would logically interpret step 3. of 1-AOI-24 as directing exactly this. Step 3 of 1-AOI-24 reads: TRIP the reactor, and GO TO E-0 Reactor Trip or Safety Injection, WHILE continuing with this instruction.
D. Incorrect: As seen in ARI-95-101, Reactor Coolant Pumps, the setpoint for annunciator window 100-D is 4.8 gpm. Plant Operation has shown that leakoff values of approximately 2.3 gpm (this value is slightly variable) are normal. The question gives the fact that leakoff for the #1 seal for the #1 RCP is 5.2 gpm. Using Attachment 1 of 1-AOI-24, one may observe that the normal operating range for the
              #1 seal leakoff is between 1 to 5 gpm when the Unit is at normal operating pressure of 2235 psig. Therefore, it is fact that #1 seal leakoff is high. The stem of the question also gives the fact that 1-TI-62-3; RCP 1 LWR BRG TEMP is at 230°F.
This is in excess of immediate trip criteria of 225°F (contained in Attachment 2 of 1-AOI-24). Two correct procedural avenues could be used to address this issue. Both will result in the same outcome. Firstly, the Unit Supervisor could immediately identify that RCP immediate trip criteria are met through memorization of the criteria of Attachment 2, RCP Immediate Shutdown Criteria and thus upon entry into 1-AOI-24 would select subsection 3.2, RCP Tripped or Shutdown Required. If the Unit Supervisor did not immediately identify that a trip of the RCP were required he would select section 3.3, #1 Seal Leakoff Flow High. Section 3.3 contains a decision step which selects whether a controlled shutdown (given that the unit is at 28% power a shutdown conducted per 1-GO-5 would be appropriate) is initially appropriate. If #1 seal leakoff is greater than or equal to 6.0 gpm then an initial shutdown is performed. If the seal leakoff is not in excess of this value, then the Unit
 
Supervisor would assign an operator to REFER TO Attachment 2 and thus utilize the Attachment to monitor for further degradation of the RCP seal package. If the Unit Supervisor did not recognize that an immediate trip of the RCP was required and because he had already bypassed the opportunity for plant shutdown (because seal leakoff were less than 6.0 gpm) then he would continue performing section 3.3 and thus maintain the Unit at power. The foregoing defends the plausibility of this distractor as this would be a correct procedural flowpath if the applicant did not understand that a trip setpoint were exceeded.
 
Question Number:       76 Tier:   1   Group:       1 K/A:   015/017 Reactor Coolant Pump (RCP) Malfunctions 2.2 Equipment Control 2.2.44 Ability to interpret control room indications to verify the status and operation ofl a system, and understand how operator actions and directives affect plant and system conditions.
Importance Rating:       4.2 4.4 10 CFR Part 55:       (CFR: 41.5 / 43.5 / 45.12) 10CFR55.43.b:         10 CFR 55.43(b)(5)
K/A Match:   K/A is matched because the applicant is required to interpret the #1 RCP parameters to verify that the RCP is operating improperly and thus that the correct operator directives will cause the plant to be tripped and the RCP to be secured in accordance with the guidance of 1-AOI-24.
Technical  
Technical  


==Reference:==
==Reference:==
ARI-95-101, Reactor Coolant Pumps 1-AOI-24, RCP Malfunctions During Pump Operation 0-TI-12.04, User's Guide for Abnormal and Emergency Operating Instructions Proposed references to be provided: None  Learning Objective: 3-OT-AOI2400 5. Given a set of plant conditions, DESCRIBE operator actions required in response to the following per AOI-24, RCP Malfunctions during Pump Operation: a. RCP tripped or shutdown required  
ARI-95-101, Reactor Coolant Pumps 1-AOI-24, RCP Malfunctions During Pump Operation 0-TI-12.04, Users Guide for Abnormal and Emergency Operating Instructions Proposed references to       None be provided:
: b. #1 Seal Leakoff Flow HIGH  
Learning Objective:         3-OT-AOI2400
: c. #1 Seal Leakoff Flow LOW AND Standpipe level alarm DARK,  
: 5. Given a set of plant conditions, DESCRIBE operator actions required in response to the following per AOI-24, RCP Malfunctions during Pump Operation:
: d. #2 Seal Leakoff Flow HIGH e. #3 Seal Leakoff Flow HIGH Cognitive Level:     Higher X Lower     Question Source:     New X Modified Bank   Bank     Question History: New question for the 2015-301 NRC SRO Exam Comments: The question is SRO only as this question meets the general SRO only criteria of Assessment of facility conditions an selection of appropriate procedures during normal, abnormal, and emergency situations.
: a. RCP tripped or shutdown required
WBN Unit 1 RCP MALFUNCTIONS DURING PUMP OPERATION 1-AOI-24 Rev. 0000 Page 5 of 27  3.0 OPERATOR ACTION: 3.1 Diagnostics IF GO TO Subsection RCP tripped or shutdown required 3.2 #1 seal leakoff flow HIGH, 3.3 #1 seal leakoff flow LOW, AND Standpipe level alarm DARK, 3.4 #2 Seal Leakoff Flow HIGH
: b. #1 Seal Leakoff Flow HIGH
(#1 seal leakoff flow LOW, AND Standpipe level alarm LIT), 3.5 #3 seal leakoff flow HIGH
: c. #1 Seal Leakoff Flow LOW AND Standpipe level alarm DARK,
(#1 seal leakoff flow NORMAL AND Standpipe level alarm LIT), 3.6 WBN Unit 1 RCP MALFUNCTIONS DURING PUMP OPERATION 1-AOI-24 Rev. 0000  Step    Action/Expected Response   Response Not Obtained   Page 11 of 27  3.3 # 1 Seal Leakoff Flow High   CAUTION A seal leakoff rise to greater than 2.0 gpm AFTER experiencing low leakoff of less than 0.8 gpm may indicate seal degradation. Plant Management should be notified of leakoff trends. NOTE 1 Anytime #1 seal leakoff flow exceeds the values shown on Attachment 1, system engineering should be requested to perform an evaluation of the #1 seal condition. NOTE 2 During plant startup after seal maintenance, the #1 seal may require 24 hours of run time before the seal seats fully and operates normally. NOTE 3 The #1 seal return should be isolated between 3 and 5 minutes after tripping an RCP to allow for pump coastdown. 1. MONITOR #1 seal leakoff equal to or greater than 6.0 gpm. **GO TO Step 5. 2. MONITOR RCPs lower bearing and #1 seal outlet temp STABLE or DROPPING. **GO TO Subsection 3.2, Step 2. 3. REFER TO appropriate instruction to initiate a controlled shutdown to Mode 3 while continuing with this instruction:
: d. #2 Seal Leakoff Flow HIGH
: e. #3 Seal Leakoff Flow HIGH Cognitive Level:
Higher               X Lower Question Source:
New                   X Modified Bank Bank Question History:           New question for the 2015-301 NRC SRO Exam Comments:                   The question is SRO only as this question meets the general SRO only criteria of Assessment of facility conditions an selection of appropriate procedures during normal, abnormal, and emergency situations.
 
WBN           RCP MALFUNCTIONS DURING PUMP           1-AOI-24 Unit 1                      OPERATION                Rev. 0000 3.0   OPERATOR ACTION:
3.1   Diagnostics IF                           GO TO Subsection RCP tripped or shutdown required                                 3.2
    #1 seal leakoff flow HIGH,                                       3.3
    #1 seal leakoff flow LOW,                                       3.4 AND                               Seal leakoff is high (at 5.2 gpm). It is Standpipe level alarm DARK,       plausible to enter this section if one
    #2 Seal Leakoff Flow HIGH                                       3.5 did not (upon entry into this procedure)
(#1 seal leakoff flow LOW,       recognize that an immediate AND                               shutdown were Standpipe level alarm LIT),       required.
    #3 seal leakoff flow HIGH                                       3.6
(#1 seal leakoff flow NORMAL AND Standpipe level alarm LIT),
Page 5 of 27
 
WBN             RCP MALFUNCTIONS DURING PUMP                 1-AOI-24 Unit 1                        OPERATION                    Rev. 0000 Step   Action/Expected Response                     Response Not Obtained 3.3     # 1 Seal Leakoff Flow High CAUTION         A seal leakoff rise to greater than 2.0 gpm AFTER experiencing low leakoff of less than 0.8 gpm may indicate seal degradation.
Plant Management should be notified of leakoff trends.
NOTE 1           Anytime #1 seal leakoff flow exceeds the values shown on Attachment 1, system engineering should be requested to perform an evaluation of the #1 seal condition.
NOTE 2           During plant startup after seal maintenance, the #1 seal may require 24 hours of run time before the seal seats fully and operates normally.
NOTE 3           The #1 seal return should be isolated between 3 and 5 minutes after tripping an RCP to allow for pump coastdown.
: 1.       MONITOR #1 seal leakoff equal to or         **GO TO Step 5.
greater than 6.0 gpm.
: 2.       MONITOR RCPs lower bearing and               **GO TO Subsection 3.2, Step 2.
        #1 seal outlet temp STABLE or DROPPING.
: 3.       REFER TO appropriate instruction to initiate a controlled shutdown to Mode 3 while continuing with this instruction:
* AOI-39, Rapid Load Reduction.
* AOI-39, Rapid Load Reduction.
* GO-4, Normal Power Operation.
* GO-4, Normal Power Operation.
* GO-5, Unit Shutdown From 30% Reactor Power to Hot Standby.
* GO-5, Unit Shutdown From 30%
WBN Unit 1 RCP MALFUNCTIONS DURING PUMP OPERATION 1-AOI-24 Rev. 0000  Step    Action/Expected Response   Response Not Obtained   3.3 # 1 Seal Leakoff Flow High (continued)   Page 12 of 27  NOTE RCP shutdown time is based on an orderly reactor shutdown and may be delayed or expedited based on ongoing evaluations of current plant conditions, other pump parameters and efforts to restore seal leakoff flows to normal. 4. REMOVE RCP from service:
Reactor Power to Hot Standby.
Page 11 of 27
 
WBN           RCP MALFUNCTIONS DURING PUMP               1-AOI-24 Unit 1                        OPERATION                  Rev. 0000 Step   Action/Expected Response                     Response Not Obtained 3.3     # 1 Seal Leakoff Flow High (continued)
NOTE           RCP shutdown time is based on an orderly reactor shutdown and may be delayed or expedited based on ongoing evaluations of current plant conditions, other pump parameters and efforts to restore seal leakoff flows to normal.
: 4.       REMOVE RCP from service:
* Within 8 hrs, OR
* Within 8 hrs, OR
* As directed by Plant Management. 5. MONITOR RCP immediate shutdown required:
* As directed by Plant Management.
* REFER TO ATTACHMENT 2, RCP Immediate Shutdown Criteria. ** GO TO Subsection 3.2, Step 2. **  GO TO Step 6. 6. ADJUST seal injection flow to exceed total #1 seal leakoff rate. 7. CONTACT System Engineer for further guidance WHILE continuing this Instruction:
: 5.       MONITOR RCP immediate                     ** GO TO Step 6.
shutdown required:
* REFER TO ATTACHMENT 2, RCP Immediate Shutdown Criteria.
        ** GO TO Subsection 3.2, Step 2.
: 6.       ADJUST seal injection flow to exceed total #1 seal leakoff rate.
: 7.       CONTACT System Engineer for further guidance WHILE continuing this Instruction:
* Recommendations for continued RCP operation.
* Recommendations for continued RCP operation.
* Installation of alternate flow measuring equipment (flows greater than 6 gpm).
* Installation of alternate flow measuring equipment (flows greater than 6 gpm).
WBN Unit 1 RCP MALFUNCTIONS DURING PUMP OPERATION 1-AOI-24 Rev. 0000  Step    Action/Expected Response   Response Not Obtained   3.3 # 1 Seal Leakoff Flow High (continued)   Page 13 of 27  CAUTION If all RCP seal cooling is lost, cooling down and depressurizing the RCS at a rapid rate, within established guidelines will minimize seal leakage. 8. CHECK seal injection flow between 8 and 13 gpm/RCP. ADJUST 1-HIC-62-89A and 1-HIC-62-93A to establish seal injection flow between 8 and 13 gpm/RCP. IF seal injection remains less than 8 gpm/RCP, THEN: a. ENSURE CCS flow to thermal barrier. b. ENSURE RCP pump lower bearing and #1 seal outlet remains less than 225°F. c. EVALUATE changing seal injection filter(s). 9. CONTROL VCT outlet temp less than 123°F:
Page 12 of 27
 
WBN           RCP MALFUNCTIONS DURING PUMP                   1-AOI-24 Unit 1                        OPERATION                        Rev. 0000 Step   Action/Expected Response                     Response Not Obtained 3.3     # 1 Seal Leakoff Flow High (continued)
CAUTION       If all RCP seal cooling is lost, cooling down and depressurizing the RCS at a rapid rate, within established guidelines will minimize seal leakage.
: 8.       CHECK seal injection flow between           ADJUST 1-HIC-62-89A and 1-HIC-62-93A 8 and 13 gpm/RCP.                          to establish seal injection flow between 8 and 13 gpm/RCP.
IF seal injection remains less than 8 gpm/RCP, THEN:
: a. ENSURE CCS flow to thermal barrier.
: b. ENSURE RCP pump lower bearing and #1 seal outlet remains less than 225°F.
: c. EVALUATE changing seal injection filter(s).
: 9.       CONTROL VCT outlet temp less than 123°F:
* ADJUST 1-HIC-62-78A.
* ADJUST 1-HIC-62-78A.
* ADJUST charging and letdown flow to reduce regenerative heat-exchanger outlet temp. 10. CHECK VCT pressure between 15 and 30 psig. ADJUST VCT pressure:
* ADJUST charging and letdown flow to reduce regenerative heat-exchanger outlet temp.
: 10.     CHECK VCT pressure between 15               ADJUST VCT pressure:
and 30 psig.
* VENT VCT by controlling 1-FCV-62-125, OR
* VENT VCT by controlling 1-FCV-62-125, OR
* CONTROL VCT level by diversion or makeup.
* CONTROL VCT level by diversion or makeup.
WBN Unit 1 RCP MALFUNCTIONS DURING PUMP OPERATION 1-AOI-24 Rev. 0000  Step    Action/Expected Response   Response Not Obtained   3.3 # 1 Seal Leakoff Flow High (continued)   Page 14 of 27  11. MONITOR RCP lower bearing and #1 seal outlet temp:
Page 13 of 27
* Less than or equal to 180°F
 
* STABLE or DROPPING. IF temp greater than 180°F AND rising, THEN **  GO TO Subsection 3.2. 12. INITIATE repairs as required. 13. RETURN TO Instruction in effect. End of Section WBN Unit 1 RCP MALFUNCTIONS DURING PUMP OPERATION 1-AOI-24 Rev. 0000 Page 26 of 27  Attachment 2 (Page 1 of 1) RCP IMMEDIATE SHUTDOWN CRITERIA   NOTE Exceeding any of the following setpoints will require an immediate pump shutdown. Operating limits can be found in SOI 68.02. This list is immediate shutdown criteria only. A. Shaft vibration greater than 20 mils or 15 mils with a rate of rise equal to 1 mil/hr (alarm at 15 mils). [Indicators located on 0-PNL-52-R139, Aux Inst Rm.] B. Frame vibration greater than 5 mils or 3 mils with a rate of rise of 0.2 mil/hr. [Readings taken by Maint. at Aux Bldg L-Panels, el.737.] C. Motor windings temp greater than 302°F. D. Motor bearing temp greater than 195°F. E. Pump bearing temp greater than 225°F. F. Loss of CCS to oil coolers for greater than 10 minutes. G. No. 1 seal outlet temp greater than 225°F. H. No. 1 seal flow HIGH with rising pump bearing or #1 seal leakoff temperatures. I. No. 1 seal P less than or equal to 200 psid.
WBN           RCP MALFUNCTIONS DURING PUMP             1-AOI-24 Unit 1                      OPERATION                  Rev. 0000 Step   Action/Expected Response                 Response Not Obtained 3.3     # 1 Seal Leakoff Flow High (continued)
WBN Unit 1 & 2 User's Guide for Abnormal and Emergency  Operating Instructions 0-TI-12.04 Rev. 0000 Page 35 of 57 2.7 Prudent Operator Actions (continued)     3. The operator should consult nearby personnel who are suitably qualified and notify them of their proposed actions. If no disagreement is forthcoming, he should then take the necessary mitigation or preemptive actions to terminate the event. 4. The STAR principle should be applied --Stop, Think, Act, Review. Ask yourself: If I take this action, could I inadvertently cause other more severe problems? Am I better off taking no action at all? How will safety status be affected? 2.8 Use of AOIs While in EOIs 1. During performance of the 1-ES-0.1, if plant conditions warrant implementation of an AOI, then the required AOI may be performed concurrently (on a not-to-interfere basis) with the EOIs. 2. When running an AOI concurrently with an EOI (1-ECA-0.0, 1-ES-0.1, etc.) the Unit Supervisor/SRO will assign the BOP/CRO operator responsibility for the AOI if another Unit Supervisor is NOT available. If the BOP/CRO operator performs an AOI, he/she should consult directly with the Unit Supervisor and give them the status as required by the AOI. 3. When an AOI in effect directs a Reactor Trip, then the performance of the AOI should continue immediately following transition to 1-ES-0.1. Performance assignments will be at the discretion of the SM/US based on the status and importance of events in progress. 4. When implementing an AOI outside the "horseshoe" in the control room, the Unit Supervisor should accompany the board operator to read the procedure steps and direct actions of the operator, unless higher priority conditions demanding the Unit Supervisor's attention exist; in which case the BOP/CRO should implement the AOI using the single performer method. The actively licensed STA may serve as a reader unless the crew is in progress of performing actions within the EOI network. 3.0 RECORDS None WBN Unit 1 RCP MALFUNCTIONS DURING PUMP OPERATION 1-AOI-24 Rev. 0000 Page 5 of 27  3.0 OPERATOR ACTION: 3.1 Diagnostics IF GO TO Subsection RCP tripped or shutdown required 3.2 #1 seal leakoff flow HIGH, 3.3 #1 seal leakoff flow LOW, AND Standpipe level alarm DARK, 3.4 #2 Seal Leakoff Flow HIGH
: 11.     MONITOR RCP lower bearing and           IF temp greater than 180°F AND rising,
(#1 seal leakoff flow LOW, AND Standpipe level alarm LIT), 3.5 #3 seal leakoff flow HIGH
        #1 seal outlet temp:                   THEN
(#1 seal leakoff flow NORMAL AND Standpipe level alarm LIT), 3.6 WBN Unit 1 RCP MALFUNCTIONS DURING PUMP OPERATION 1-AOI-24 Rev. 0000  Step    Action/Expected Response   Response Not Obtained   Page 6 of 27  3.2 RCP Tripped Or Shutdown Required   NOTE 1 Malfunctions addressed by this procedure require RCP shutdown as soon as possible. NOTE 2 Exceeding any of the limits listed on Attachment 2 of this procedure will require immediate shutdown of the affected RCP. NOTE 3 Malfunctions resulting in high #1 seal leakoff will require closing #1 seal return FCV following RCP coastdown 1. CHECK RCP tripped MONITOR RCP immediate shutdown Criteria:
* Less than or equal to 180°F       ** GO TO Subsection 3.2.
* REFER TO ATTACHMENT 2, RCP Immediate Shutdown Criteria. 1) IF RCP immediate shutdown required, THEN   ** GO TO Step 2. 2) IF RCP immediate shutdown NOT required, THEN   ** GO TO Step 9 2. CHECK unit in Mode 1 or 2 ** GO TO Step 4.
* STABLE or DROPPING.
WBN Unit 1 RCP MALFUNCTIONS DURING PUMP OPERATION 1-AOI-24 Rev. 0000  Step    Action/Expected Response   Response Not Obtained   3.2 RCP Tripped Or Shutdown Required (continued)   Page 7 of 27    NOTE Control room staff should brief on Steps 3 through 6 prior to tripping the reactor. This ensures that the affected RCP is stopped and that appropriate actions are taken when unit is removed from service. 3. TRIP the reactor, and GO TO E-0, Reactor Trip or Safety Injection, WHILE continuing with this instruction. . 4. STOP and LOCK OUT affected RCP(s). 5. IF in Mode 3, THEN CHECK any RCP Running **GO TO ES-0.2, Natural Circulation Cooldown, WHILE continuing with this instruction   CAUTION If the RCP seal return flow control valve (FCV) is NOT closed within 5 minutes of stopping the RCP with excessive leakoff, seal damage may occur. 6. MONITOR RCP seal leakoff less than 6 gpm per pump:
: 12.     INITIATE repairs as required.
: 13.     RETURN TO Instruction in effect.
End of Section Page 14 of 27
 
WBN           RCP MALFUNCTIONS DURING PUMP                 1-AOI-24 Unit 1                      OPERATION                      Rev. 0000 Attachment 2 (Page 1 of 1)
RCP IMMEDIATE SHUTDOWN CRITERIA NOTE           Exceeding any of the following setpoints will require an immediate pump shutdown. Operating limits can be found in SOI 68.02. This list is immediate shutdown criteria only.
A. Shaft vibration greater than 20 mils or 15 mils with a rate of rise equal to 1 mil/hr (alarm at 15 mils). [Indicators located on 0-PNL-52-R139, Aux Inst Rm.]
B. Frame vibration greater than 5 mils or 3 mils with a rate of rise of 0.2 mil/hr.
[Readings taken by Maint. at Aux Bldg L-Panels, el.737.]
C. Motor windings temp greater than 302°F.
D. Motor bearing temp greater than 195°F.
E. Pump bearing temp greater than 225°F.
F. Loss of CCS to oil coolers for greater than 10 minutes.
G. No. 1 seal outlet temp greater than 225°F.
H. No. 1 seal flow HIGH with rising pump bearing or #1 seal leakoff temperatures.
I. No. 1 seal P less than or equal to 200 psid.
Page 26 of 27
 
WBN                       User's Guide for                 0-TI-12.04 Unit 1 & 2          Abnormal and Emergency                  Rev. 0000 Operating Instructions                Page 35 of 57 2.7     Prudent Operator Actions (continued)
: 3. The operator should consult nearby personnel who are suitably qualified and notify them of their proposed actions. If no disagreement is forthcoming, he should then take the necessary mitigation or preemptive actions to terminate the event.
: 4. The STAR principle should be applied --Stop, Think, Act, Review. Ask yourself: If I take this action, could I inadvertently cause other more severe problems? Am I better off taking no action at all? How will safety status be affected?
2.8     Use of AOIs While in EOIs
: 1. During performance of the 1-ES-0.1, if plant conditions warrant implementation of an AOI, then the required AOI may be performed concurrently (on a not-to-interfere basis) with the EOIs.
: 2. When running an AOI concurrently with an EOI (1-ECA-0.0, 1-ES-0.1, etc.)
the Unit Supervisor/SRO will assign the BOP/CRO operator responsibility for the AOI if another Unit Supervisor is NOT available. If the BOP/CRO operator performs an AOI, he/she should consult directly with the Unit Supervisor and give them the status as required by the AOI.
: 3. When an AOI in effect directs a Reactor Trip, then the performance of the AOI should continue immediately following transition to 1-ES-0.1.
Performance assignments will be at the discretion of the SM/US based on the status and importance of events in progress.
: 4. When implementing an AOI outside the horseshoe in the control room, the Unit Supervisor should accompany the board operator to read the procedure steps and direct actions of the operator, unless higher priority conditions demanding the Unit Supervisors attention exist; in which case the BOP/CRO should implement the AOI using the single performer method. The actively licensed STA may serve as a reader unless the crew is in progress of performing actions within the EOI network.
3.0     RECORDS None
 
WBN           RCP MALFUNCTIONS DURING PUMP   1-AOI-24 Unit 1                      OPERATION        Rev. 0000 3.0   OPERATOR ACTION:
3.1   Diagnostics IF                   GO TO Subsection RCP tripped or shutdown required                       3.2
    #1 seal leakoff flow HIGH,                             3.3
    #1 seal leakoff flow LOW,                               3.4 AND Standpipe level alarm DARK,
    #2 Seal Leakoff Flow HIGH                               3.5
(#1 seal leakoff flow LOW, AND Standpipe level alarm LIT),
    #3 seal leakoff flow HIGH                               3.6
(#1 seal leakoff flow NORMAL AND Standpipe level alarm LIT),
Page 5 of 27
 
WBN         RCP MALFUNCTIONS DURING PUMP               1-AOI-24 Unit 1                      OPERATION                    Rev. 0000 Step   Action/Expected Response                 Response Not Obtained 3.2     RCP Tripped Or Shutdown Required NOTE 1       Malfunctions addressed by this procedure require RCP shutdown as soon as possible.
NOTE 2       Exceeding any of the limits listed on Attachment 2 of this procedure will require immediate shutdown of the affected RCP.
NOTE 3       Malfunctions resulting in high #1 seal leakoff will require closing #1 seal return FCV following RCP coastdown
: 1.     CHECK RCP tripped                         MONITOR RCP immediate shutdown Criteria:
* REFER TO ATTACHMENT 2, RCP Immediate Shutdown Criteria.
: 1)   IF RCP immediate shutdown required, THEN
                                                        ** GO TO Step 2.
: 2)   IF RCP immediate shutdown NOT required, THEN
                                                        ** GO TO Step 9
: 2.     CHECK unit in Mode 1 or 2                 ** GO TO Step 4.
Page 6 of 27
 
WBN           RCP MALFUNCTIONS DURING PUMP               1-AOI-24 Unit 1                      OPERATION                    Rev. 0000 Step   Action/Expected Response                   Response Not Obtained 3.2     RCP Tripped Or Shutdown Required (continued)
NOTE           Control room staff should brief on Steps 3 through 6 prior to tripping the reactor. This ensures that the affected RCP is stopped and that appropriate actions are taken when unit is removed from service.
: 3.     TRIP the reactor, and                       .
GO TO E-0, Reactor Trip or Safety Injection, WHILE continuing with this instruction.
: 4.     STOP and LOCK OUT affected RCP(s).
: 5.     IF in Mode 3,                               **GO TO ES-0.2, Natural Circulation THEN                                        Cooldown, WHILE continuing with this CHECK any RCP Running                        instruction CAUTION       If the RCP seal return flow control valve (FCV) is NOT closed within 5 minutes of stopping the RCP with excessive leakoff, seal damage may occur.
: 6.     MONITOR RCP seal leakoff less               WHEN the RCP has coasted down than 6 gpm per pump:                       (between 3 and 5 minutes),
THEN
* 1-FR-62-24 [RCP 1 & 2]
* 1-FR-62-24 [RCP 1 & 2]
* 1-FR-62-50 [RCP 3 & 4]
* 1-FR-62-50 [RCP 3 & 4]               CLOSE affected RCP seal return FCV:
* ICS "RCP DATA"
* ICS RCP DATA
* ICS "RCP SEALS" WHEN the RCP has coasted down (between 3 and 5 minutes),  THEN  CLOSE affected RCP seal return FCV:
* 1-FCV-62-9 [RCP 1]
* 1-FCV-62-9 [RCP 1]
* ICS RCP SEALS
* 1-FCV-62-22 [RCP 2]
* 1-FCV-62-22 [RCP 2]
* 1-FCV-62-35 [RCP 3]
* 1-FCV-62-35 [RCP 3]
* 1-FCV-62-48 [RCP 4] 7. CHECK RCPs 1 and 2 running. CLOSE affected loop's pressurizer spray valve.
* 1-FCV-62-48 [RCP 4]
WBN Unit 1 RCP MALFUNCTIONS DURING PUMP OPERATION 1-AOI-24 Rev. 0000  Step    Action/Expected Response   Response Not Obtained   3.2 RCP Tripped Or Shutdown Required (continued)   Page 8 of 27  8. GO TO Step 15. 9. CONSULT plant staff as necessary for recommendations for continued RCP operation. NOTE Control room staff should brief on Steps 10 through 13 prior to reducing load. This ensures that the affected RCP is stopped and that appropriate actions are taken when unit is removed from service. 10. IF removal of RCP(s) is required, THEN REFER TO appropriate instruction to initiate a controlled shutdown to Mode 3 while continuing with this instruction:
: 7.     CHECK RCPs 1 and 2 running.                 CLOSE affected loops pressurizer spray valve.
Page 7 of 27
 
WBN           RCP MALFUNCTIONS DURING PUMP               1-AOI-24 Unit 1                      OPERATION                    Rev. 0000 Step   Action/Expected Response                   Response Not Obtained 3.2     RCP Tripped Or Shutdown Required (continued)
: 8.     GO TO Step 15.
: 9.     CONSULT plant staff as necessary for recommendations for continued RCP operation.
NOTE           Control room staff should brief on Steps 10 through 13 prior to reducing load. This ensures that the affected RCP is stopped and that appropriate actions are taken when unit is removed from service.
: 10. IF removal of RCP(s) is required,         RETURN TO instruction in effect.
THEN REFER TO appropriate instruction to initiate a controlled shutdown to Mode 3 while continuing with this instruction:
* AOI-39, Rapid Load Reduction
* AOI-39, Rapid Load Reduction
* GO-4, Normal Power Operation.
* GO-4, Normal Power Operation.
* GO-5, Unit Shutdown From 30% Reactor Power to Hot Standby RETURN TO instruction in  effect. 11. MAINTAIN affected SG level on PROGRAM:
* GO-5, Unit Shutdown From 30%
Reactor Power to Hot Standby
: 11. MAINTAIN affected SG level on PROGRAM:
* LOWER MFW flow as steam flow drops.
* LOWER MFW flow as steam flow drops.
* ISOLATE blowdown from affected SG. 12. WHEN unit is in Mode 3, THEN WBN Unit 1 RCP MALFUNCTIONS DURING PUMP OPERATION 1-AOI-24 Rev. 0000  Step    Action/Expected Response   Response Not Obtained   3.2 RCP Tripped Or Shutdown Required (continued)   Page 9 of 27  a. STOP and LOCK OUT affected RCP(s). b. CHECK any RCP Running b. **GO TO ES-0.2, Natural Circulation Cooldown, WHILE continuing with this instruction. CAUTION If the RCP seal return flow control valve (FCV) is NOT closed within 5 minutes of stopping the RCP with excessive leakoff, seal damage may occur. 13. MONITOR RCP seal leakoff less than 6 gpm per pump:
* ISOLATE blowdown from affected SG.
: 12. WHEN unit is in Mode 3, THEN Page 8 of 27
 
WBN         RCP MALFUNCTIONS DURING PUMP                 1-AOI-24 Unit 1                    OPERATION                      Rev. 0000 Step   Action/Expected Response                 Response Not Obtained 3.2     RCP Tripped Or Shutdown Required (continued)
: a. STOP and LOCK OUT affected RCP(s).
: b. CHECK any RCP Running               b.   **GO TO ES-0.2, Natural Circulation Cooldown, WHILE continuing with this instruction.
CAUTION       If the RCP seal return flow control valve (FCV) is NOT closed within 5 minutes of stopping the RCP with excessive leakoff, seal damage may occur.
: 13. MONITOR RCP seal leakoff less than       WHEN the RCP has coasted down 6 gpm per pump:                           (between 3 and 5 minutes),
THEN
* 1-FR-62-24 [RCP 1 & 2]
* 1-FR-62-24 [RCP 1 & 2]
* 1-FR-62-50 [RCP 3 & 4] WHEN the RCP has coasted down (between 3 and 5 minutes),  THEN  CLOSE affected RCP seal return FCV:
CLOSE affected RCP seal return FCV:
* 1-FR-62-50 [RCP 3 & 4]
* 1-FCV-62-9 [RCP 1]
* 1-FCV-62-9 [RCP 1]
* 1-FCV-62-22 [RCP 2]
* 1-FCV-62-22 [RCP 2]
* 1-FCV-62-35 [RCP 3]
* 1-FCV-62-35 [RCP 3]
* 1-FCV-62-48 [RCP 4] 14. CHECK RCPs 1 and 2 running. CLOSE affected loop's pressurizer spray valve.
* 1-FCV-62-48 [RCP 4]
WBN Unit 1 RCP MALFUNCTIONS DURING PUMP OPERATION 1-AOI-24 Rev. 0000  Step    Action/Expected Response   Response Not Obtained   3.2 RCP Tripped Or Shutdown Required (continued)   Page 10 of 27  15. REFER TO Tech Spec:
: 14. CHECK RCPs 1 and 2 running.               CLOSE affected loops pressurizer spray valve.
Page 9 of 27
 
WBN           RCP MALFUNCTIONS DURING PUMP         1-AOI-24 Unit 1                      OPERATION                Rev. 0000 Step   Action/Expected Response                 Response Not Obtained 3.2     RCP Tripped Or Shutdown Required (continued)
: 15. REFER TO Tech Spec:
* LCO 3.4.1, RCS Press, Temp and Flow DNB Limits.
* LCO 3.4.1, RCS Press, Temp and Flow DNB Limits.
* LCO 3.4.2, RCS Minimum Temp For Criticality.
* LCO 3.4.2, RCS Minimum Temp For Criticality.
Line 80: Line 230:
* LCO 3.4.5, RCS Loops Mode 3.
* LCO 3.4.5, RCS Loops Mode 3.
* LCO 3.4.13, RCS Operational Leakage.
* LCO 3.4.13, RCS Operational Leakage.
* LCO 3.4.15, Leakage Detection Instrumentation. 16. INITIATE repairs as required. 17. OBTAIN plant management approval prior to restarting any RCP. 18. RETURN TO Instruction in effect. End of Section WBN Unit 1 Reactor Coolant Pumps ARI-95-101 Rev. 0035 Page 39 of 50   A. No. 1 seal damage B. No. 1 seal NOT fully seated C. Loss of seal injection water followed by high seal temperature [1] VERIFY high leakoff flow condition of affected RCP(s) with the following instruments: RCP RECORDER PEN/TRACE ICS POINT 1 1-FR-62-24 Red F1018A 2 1-FR-62-24 Blue F1020A 3 1-FR-62-50 Red F1022A 4 1-FR-62-50 Blue F1024A [2] IF high leakoff is confirmed, THEN GO TO AOI-24, RCP MALFUNCTIONS DURING PUMP OPERATION. 1-47W610-62-1 AOI-24  Source Setpoint RCP 1: 1-FS-62-11 4.8 gpm RCP 2: 1-FS-62-24  RCP 3: 1-FS-62-37  RCP 4: 1-FS-62-50        100-DRCP SEAL LEAK OFF FLOW HI (Page 1 of 1) Probable Cause: Corrective Action:
* LCO 3.4.15, Leakage Detection Instrumentation.
: 16. INITIATE repairs as required.
: 17. OBTAIN plant management approval prior to restarting any RCP.
: 18. RETURN TO Instruction in effect.
End of Section Page 10 of 27
 
WBN                   Reactor Coolant Pumps                 ARI-95-101 Unit 1                                                        Rev. 0035 Page 39 of 50 100-D Source                            Setpoint RCP 1: 1-FS-62-11                  4.8 gpm                                      RCP SEAL LEAK OFF RCP 2: 1-FS-62-24 FLOW RCP 3: 1-FS-62-37                                                                  HI RCP 4: 1-FS-62-50 (Page 1 of 1)
Probable        A. No. 1 seal damage Cause:          B. No. 1 seal NOT fully seated C. Loss of seal injection water followed by high seal temperature Corrective      [1] VERIFY high leakoff flow condition of affected RCP(s) with the following Action:            instruments:
RCP           RECORDER               PEN/TRACE               ICS POINT 1           1-FR-62-24                 Red                 F1018A 2           1-FR-62-24                 Blue                 F1020A 3           1-FR-62-50                 Red                 F1022A 4           1-FR-62-50                 Blue                 F1024A
[2] IF high leakoff is confirmed, THEN GO TO AOI-24, RCP MALFUNCTIONS DURING PUMP OPERATION.


==References:==
==References:==
1-47W610-62-1 AOI-24


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 6 of 16  C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)] Some examples of SRO exam items for this topic include:
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]
Some examples of SRO exam items for this topic include:
* 10 CFR 50.59 screening and evaluation processes.
* 10 CFR 50.59 screening and evaluation processes.
* Administrative processes for temporary modifications.
* Administrative processes for temporary modifications.
* Administrative processes for disabling annunciators.
* Administrative processes for disabling annunciators.
* Administrative processes for the installation of temporary instrumentation.
* Administrative processes for the installation of temporary instrumentation.
* Processes for changing the plant or plant procedures. Section IV provides an example of a satisfactory SRO-only question related to this topic. D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.       [10 CFR 55.43(b)(4)] Some examples of SRO exam items for this topic include:
* Processes for changing the plant or plant procedures.
Section IV provides an example of a satisfactory SRO-only question related to this topic.
D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
[10 CFR 55.43(b)(4)]
Some examples of SRO exam items for this topic include:
* Process for gaseous/liquid release approvals, i.e., release permits.
* Process for gaseous/liquid release approvals, i.e., release permits.
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures. Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits. SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc. E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedure's overall mitigative strategy or purpose.  
Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.
 
SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.
The applicant's knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedure's content is required to correctly answer the written test item, for example:
E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.
The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:
Page 6 of 16


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 7 of 16
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
* Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
* Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
SRO-only knowledge should not be claimed for questions that can be answered solely using "systems knowledge"; e.g.:
SRO-only knowledge should not be claimed for questions that can be answered solely using systems knowledge; e.g.:
* how the system works.
* how the system works.
* system flow path.
* system flow path.
* component locations, etc. SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:
* component locations, etc.
SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:
* the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
* the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
* any AOP entry condition.
* any AOP entry condition.
* plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
* plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
* immediate operator actions of a procedure. Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.        
* immediate operator actions of a procedure.
Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.
Page 7 of 16
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)
(Assessment and selection of procedures)
Can the question be answered solely by knowing systems knowledge, i.e., how the system works,      Yes RO question flowpath, logic, component location?
No Can the question be answered solely by knowing immediate operator actions?                            Yes    RO question No Can the question be answered solely by knowing entry conditions for AOPs or plant parameters          Yes RO question that require direct entry to major EOPs?
No Can the question be answered solely by knowing the purpose, overall sequence of events, or            Yes RO question overall mitigative strategy of a procedure?
No Does the question require one or more of the following?
* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps                                            Yes        SRO-only
* Knowledge of diagnostic steps and decision points in the              question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 8 of 16   Figure 2:  Screening for SRO-only linked to 10 CFR 55.43(b)(5) (Assessment and selection of procedures)      
77.
Given the following conditions:
    -  1-FR-S.1, Nuclear Power Generation/ATWS was entered.
    -  The Unit Supervisor has reached step 12, MONITOR reactor subcriticality:
    -  The following indications are noted:
Which ONE of the following completes the statements listed below?
In accordance with 1-FR-S.1, the conditions shown above _____(1)_____ CURRENTLY allow emergency boration to be terminated.
In accordance with the Westinghouse background document for 1-FR-S.1, Step 4, INITIATE RCS Boration: _____(2)_____ a TIME CRITICAL step.
A.    (1)  does (2)  is B.    (1)  does (2)  is NOT C.     (1)   does NOT (2)   is D.    (1)  does NOT (2)   is NOT


Can the question be answered solely by knowing "systems knowledge", i.e., how the system works, flowpath, logic, component location? RO question YesNoCan the question be answered solely by knowing immediate operator actions? YesCan the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?YesNoDoes the question require one or more of the following?
CORRECT ANSWER:                                                                           D DISTRACTOR ANALYSIS:
* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
A. Incorrect: Step 12 of 1-FR-S.1 states: MONITOR reactor subcriticality: a. CHECK Power range channels less than 5%. b. CHECK Intermediate range startup rate NEGATIVE. The conditions displayed in the stem of the question indicate that the PRNIs are at 4% but that the IR SUR is 0 dpm (i.e. not negative). Therefore, the procedure user is directed to continue in 1-FR-S.1 and not terminate emergency boration. It is plausible to believe that an IR SUR of 0 would allow the termination of emergency boration because the status tree for subcriticality allows for a ZERO SUR as a check for reactor subcriticality (in the decision tree INTERMEDIATE RANGE SUR ZERO OR NEGATIVE).
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps
While the Westinghouse background document for 1-FR-S.1 states that Emergency Boration of the RCS is the most direct manner of adding negative reactivity to the core, it does not regard this step as time critical. The foregoing supports the plausibility for the belief that Emergency Boration is time critical.
* Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
B. Incorrect: Again, it is incorrect and yet plausible that the conditions shown in the stem of the question do allow emergency boration to be terminated. Also, it is correct that the initiation of Emergency Boration is not a time critical step.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?NoNoYesQuestion might not be linked to 10 CFR 55.43(b)(5) for SRO-only NoRO question RO question RO question 
C. Incorrect: It is correct that the conditions displayed do NOT allow the boration to be terminated. It is incorrect and yet plausible that the initiation of boration is a time critical step.
: 77. Given the following conditions: FR-S.1, Nuclear Power Generation/ATWS was entered. - The Unit Supervisor has reached step 12, MONITOR reactor subcriticality: - The following indications are noted:  Which ONE of the following completes the statements listed below? In accordance with 1-FR-S.1, the conditions shown above _____(1)_____ CURRENTLY allow emergency boration to be terminated. In accordance with the Westinghouse background document for 1-FR-S.1, Step 4, INITIATE RCS Boration: _____(2)_____ a TIME CRITICAL step. A. (1) does (2) is B. (1) does  (2) is NOT C. (1) does NOT (2) is D. (1) does NOT (2) is NOT CORRECT ANSWER:D DISTRACTOR ANALYSIS:   A. Incorrect: Step 12 of 1-FR-S.1 states: "MONITOR reactor subcriticality: a. CHECK Power range channels less than 5%. b. CHECK Intermediate range startup rate NEGATIVE."  The conditions displayed in the stem of the question indicate that the PRNIs are at 4% but that the IR SUR is 0 dpm (i.e. not negative). Therefore, the procedure user is directed to continue in 1-FR-S.1 and not terminate emergency boration. It is plausible to believe that an IR SUR of 0 would allow the termination of emergency boration because the status tree for subcriticality allows for a ZERO SUR as a check for reactor subcriticality (in the decision tree "INTERMEDIATE RANGE SUR ZERO OR NEGATIVE).
D. Correct: It is correct that the conditions displayed do NOT allow the boration to be terminated. It is correct that the initiation of boration is NOT a time critical step.
While the Westinghouse background document for 1-FR-S.1 states that Emergency Boration of the RCS is "the most direct manner of adding negative reactivity to the core," it does not regard this step as time critical. The foregoing supports the plausibility for the belief that Emergency Boration is time critical. B. Incorrect: Again, it is incorrect and yet plausible that the conditions shown in the stem of the question do allow emergency boration to be terminated. Also, it is correct that the initiation of Emergency Boration is not a time critical step. C. Incorrect: It is correct that the conditions displayed do NOT allow the boration to be terminated. It is incorrect and yet plausible that the initiation of boration is a time critical step. D. Correct: It is correct that the conditions displayed do NOT allow the boration to be terminated. It is correct that the initiation of boration is NOT a time critical step.
Question Number: 77  Tier:  1  Group:  1 K/A: 029 Anticipated Transient Without Scram (ATWS) EA2 Ability to determine or interpret the following as they apply to a ATWS:
EA2.01 Reactor nuclear instrumentation  Importance Rating: 4.4  4.7


10 CFR Part 55: (CFR 43.5 / 45.13) 10CFR55.43.b: 10 CFR 55.43(b)(5) K/A Match: K/A is matched because the applicant is required to interpret both the power range NI readings as well as the intermediate range startup rate indications as they apply to an ATWS. The applicant must then decide if the sub criticality criteria are met and the subsequent required procedural actions. Technical  
Question Number:        77 Tier:    1    Group:      1 K/A:    029 Anticipated Transient Without Scram (ATWS)
EA2 Ability to determine or interpret the following as they apply to a ATWS:
EA2.01 Reactor nuclear instrumentation Importance Rating:      4.4 4.7 10 CFR Part 55:       (CFR 43.5 / 45.13) 10CFR55.43.b:         10 CFR 55.43(b)(5)
K/A Match:   K/A is matched because the applicant is required to interpret both the power range NI readings as well as the intermediate range startup rate indications as they apply to an ATWS. The applicant must then decide if the sub criticality criteria are met and the subsequent required procedural actions.
Westinghouse Background Document for 1-FR-S.1 Technical  


==Reference:==
==Reference:==
Westinghouse Background Document for 1-FR-S.1 1-FR-S.1, Nuclear Power Generation/ATWS FR-0, Status Trees Proposed references to be provided: None  Learning Objective: 3-OT-FRS0001 9. Given a set of plant conditions, use 1-FR-S.1, FR-S.2 and the Critical Safety Function Status Trees to correctly DIAGNOSE and implement:
1-FR-S.1, Nuclear Power Generation/ATWS FR-0, Status Trees Proposed references to     None be provided:
Action Steps, RNOs, Notes and Cautions. 10. EXPLAIN the purpose for and basis of each step in 1-FR-S.1 and FR-S.2 Cognitive Level:     Higher X Lower     Question Source:     New X Modified Bank   Bank     Question History: New question for the 2015-301 NRC SRO Exam Comments: The question is SRO only as this question meets the general SRO only criteria of "Assessment of facility conditions an selection of appropriate procedures during normal, abnormal, and emergency situations."
Learning Objective:       3-OT-FRS0001
WBN Unit 1 Status Trees FR-0 Rev. 0014 Page 4 of 11  Attachment 1 (Page 1 of 8) Monitoring Critical Safety Functions SUBCRITICALITY FR-S WBN Unit 1 Nuclear Power Generation/ATWS 1-FR-S.1 Rev. 0001   Step    Action/Expected Response   Response Not Obtained   Page 6 of 16  9. ENSURE the following trips: a. Reactor Trip. a. DISPATCH operator to locally trip reactor:
: 9. Given a set of plant conditions, use 1-FR-S.1, FR-S.2 and the Critical Safety Function Status Trees to correctly DIAGNOSE and implement:
* OPEN reactor trip breakers and MG set output breakers [MG set room].
Action Steps, RNOs, Notes and Cautions.
* OPEN breakers to MG sets [480V unit boards A and B]. b. Turbine Trip. b. DISPATCH operator to locally trip turbine:
: 10. EXPLAIN the purpose for and basis of each step in 1-FR-S.1 and FR-S.2 Cognitive Level:
Higher             X Lower Question Source:
New                 X Modified Bank Bank Question History:         New question for the 2015-301 NRC SRO Exam Comments:                 The question is SRO only as this question meets the general SRO only criteria of Assessment of facility conditions an selection of appropriate procedures during normal, abnormal, and emergency situations.
 
WBN         Status Trees                 FR-0 Unit 1                                  Rev. 0014 Attachment 1 (Page 1 of 8)
Monitoring Critical Safety Functions SUBCRITICALITY FR-S Page 4 of 11
 
WBN               Nuclear Power Generation/ATWS       1-FR-S.1 Unit 1                                                  Rev. 0001 Step    Action/Expected Response                 Response Not Obtained
: 9.       ENSURE the following trips:
: a. Reactor Trip.                     a. DISPATCH operator to locally trip reactor:
* OPEN reactor trip breakers and MG set output breakers
[MG set room].
* OPEN breakers to MG sets
[480V unit boards A and B].
: b. Turbine Trip.                     b. DISPATCH operator to locally trip turbine:
* TRIP from front standard.
* TRIP from front standard.
* STOP and PULL TO LOCK both EHC pumps. 10. MAINTAIN rod insertion UNTIL rods fully inserted. 11. REFER TO EPIP-1, Emergency Plan Classification Flowchart for ATWS event. 12. MONITOR reactor subcriticality: a. CHECK Power range channels less than 5%. a. ** GO TO Step 13. b. CHECK Intermediate range startup rate NEGATIVE. b. ** GO TO Step 13. c. ** GO TO Step 21.
* STOP and PULL TO LOCK both EHC pumps.
WBN Unit 1 Nuclear Power Generation/ATWS 1-FR-S.1 Rev. 0001  Step    Action/Expected Response   Response Not Obtained   Page 10 of 16  19. CHECK Incore T/Cs less than 1200°F. IF Incore T/Cs are greater than 1200°F AND rising, THEN ** GO TO 1-SACRG-1, Severe Accident Control Room Guideline Initial Response. 20. CHECK reactor subcritical: a. Power range channels less than 5%. b. Intermediate range startup rate NEGATIVE. CONTINUE to borate. IF boration is NOT available, THEN ALLOW RCS to heat up to insert negative reactivity from temperature coefficients. IF red OR orange condition exists on other Status Trees, THEN PERFORM actions of other FR Procedures which do not cool down or otherwise add positive reactivity to the core. ** GO TO Step 4. 21. TERMINATE emergency boration: a. PLACE BA transfer pumps in SLOW speed. b. CLOSE emergency borate valve 1-FCV-62-138. c. IF alternate boration opened, THEN Locally CLOSE 1-ISV-62-929.
: 10.       MAINTAIN rod insertion UNTIL rods fully inserted.
STEPDESCRIPTIONTABLEFORFR-S.lSTEP:VerifyTurbineTripPURPOSE:ToensurethattheturbineistrippedBASIS:Step__2__TheturbineistrippedtopreventanuncontrolledcooldownoftheRCSduetosteamflowthattheturbinewouldrequire.ForanATWSeventwherealossofnormalfeedwaterhasoccurred,analyseshaveshownthataturbinetripisnecessary(within30seconds)tomaintainSGinventory.Iftheturbinewillnottrip,aturbinerunback(manualdecreaseinload)atmaximumratewillalsoreducesteamflowinadelayedmanner.Iftheturbinestopvalvescannotbeclosedbyeithertriporrunback,theMSIVsshouldbeclosed.ACTIONS:oDetermineifallturbinestopvalvesareclosedoDetermineifturbinewillnottripoDetermineifturbinecannotberunbackoTriptheturbineoManuallyrunbackturbineoClosemainsteamlineisolationandbypassvalvesINSTRUMENTATION:oTurbinestopvalvepositionindicationoMSIVsandbypassvalvespositionindicationCONTROL/EQUIPMENT:oSwitchesforturbinetrip(e.g.manualtripbuttons,overspeedtestswitch,EHcontroloilpumpswitches)oControlstomanuallyrunbackturbineoSwitchestocloseMSIVsandbypassvalvesFR-S.lBackgroundHFRSIBG.doc77HP-Rev.2,4/30/2005 STEPDESCRIPTIONTABLEFORFR-S.lSTEP:CheckAFWPumpsRunningPURPOSE:ToensureAFWpumpsarerunningBASIS:Step3__TheMDAFWpumpsstartautomaticallyonanSIsignalandSGlowleveltoprovidefeedtotheSGsfordecayheatremoval.IfSGlevelsdropbelowtheappropriatesetpoint,theturbine-drivenAFWpumpwillalsoautomaticallystarttosupplementtheMDpumps.TheATWSanalyseshaveshownthatactuationofAFWwithin60secondsafterthefailuretoscramprovidesacceptableresults.ACTIONS:oDetermineifMDAFWpumpsarerunningoDetermineiftheturbine-drivenAFWpumpisrunningifnecessaryoStartMDAFWpumpsoOpensteamsupplyvalvestoturbine-drivenAFWpumpINSTRUMENTATION:oMDAFWpumpsstatusindicationoTurbine-drivenAFWpumpstatusindicationoTurbine-drivenAFWpumpsteamsupplyvalvepositionindicationCONTROL/EQUIPMENT:Switchesfor:oMDAFWpumpsoTurbine-drivenAFWpumpsteamsupplyvalvesKNOWLEDGE:N/APLANT-SPECIFICINFORMATION:N/AFR-S.lBackgroundHFRSIBG.doc79HP-Rev.2,4/30/2005 STEPDESCRIPTIONTABLEFORFR-S.lSTEP:InitiateEmergencyBorationofRCSStep4__PURPOSE:ToaddnegativereactivitytobringthereactorcoresubcriticalBASIS:Aftercontrolrodtripandrodinsertionfunctions,borationisthenextmostdirectmannerofaddingnegativereactivitytothecore.Theintendedborationpathhereisthemostdirectoneavailable,notrequiringSIinitiation,butusingnormalchargingpump(s).PumpminiflowlinesareassumedtobeopentoprotectthepumpsintheeventofhighRCSpressure.Severalplantspecificmeansareusuallyavailableforrapidborationandshouldbespecifiedhereinorderofpreference.Methodsofrapidborationincludeemergencyboration,injectingtheBIT,andsafetyinjectionactuation.ItshouldbenotedthatSIactuationwilltripthemainfeedwaterpumps.Ifthisisundesirable,theoperatorcanmanuallyalignthesystemforsafetyinjection.However,theRWSTvalvestothesuctionoftheSIpumpsshouldbeopenedfirstbeforeopeninguptheBITvalves.Ifasafetyinjectionisalreadyinprogressbutishavingnoeffectonnuclearflux,thentheBITandRWSTarenotperformingtheirintendedfunction,perhapsduetoblockageorleakage.InthiscasesomeotheralignmentusingtheBATsand/orsafeguardschargingpump(s)isrequired.ThecheckonRCSpressureisintendedtoalerttheoperatortoaconditionwhichwouldreducechargingorSIpumpinjectionintotheRCSand,therefore,boration.ThePRZRPORVpressuresetpointischosenasthatpressureatwhichflowintotheRCSisinsufficient.Thecontingentactionisarapiddepressurizationtoapressurewhichwouldallowincreasedinjectionflow.Whenprimarypressuredrops200psibelowthePORVpressuresetpoint,thePORVsshouldbeclosed.TheoperatormustverifysuccessfulclosureofthePORVs,closingtheisolationvalves,ifnecessary.FR-S.lBackgroundHFRSIBG.doc80HP-Rev.2,4/30/2005 ACTIONS:STEPDESCRIPTIONTABLEFORFR-S.lStep4__oDetermineifPRZRpressureislessthan(A.02)psigoDetermineifPRZRPORVsandblockvalvesareopenoStartcharging/SIpumpsoStartPOpumpoAlignborationpathoAlignchargingflowpathoOpenPRZRPORVsandblockvalvesasnecessaryuntilPRZRpressureislessthan(A.08)psigINSTRUMENTATION:oCharging/SIpump(s)statusindicationoPOpumpstatusindicationoPositionindicationforchargingpathvalves,borationpathvalvesoPRZRpressureindicationoPRZRPORVandblockvalvepositionindicationsCONTROL/EQUIPMENT:oCharging/SIpump(s)switchesoPOpumpswitchoSwitchesforchargingpathvalves/borationpathvalvesoPRZRPORVsandblockvalvesswitchesKNOWLEDGE:N/APLANT-SPECIFICINFORMATION:o(A.02)PRZRPORVpressuresetpoint.o(A.08)200psilessthanPRZRPORVpressuresetpoint.oPreferredalignmentsforemergencyborationbasedonplantequipmentandoperatingpractices.FR-S.1BackgroundHFRS1BG.doc81HP-Rev.2,4/30/2005 Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 6 of 16  C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)] Some examples of SRO exam items for this topic include:
: 11.       REFER TO EPIP-1, Emergency Plan Classification Flowchart for ATWS event.
: 12.       MONITOR reactor subcriticality:
: a. CHECK Power range channels         a. ** GO TO Step 13.
less than 5%.
: b. CHECK Intermediate range           b. ** GO TO Step 13.
startup rate NEGATIVE.
: c.   ** GO TO Step 21.                               POWER MUST BE LESS THAN 5%
and SUR must be NEGATIVE.
step 21 is terminate                                                    This is different emergency boration                                                      than the status trees which list IR Page 6 of 16            SUR is 0.
 
WBN             Nuclear Power Generation/ATWS             1-FR-S.1 Unit 1                                                    Rev. 0001 Step   Action/Expected Response                 Response Not Obtained
: 19. CHECK Incore T/Cs less than               IF Incore T/Cs are greater than 1200°F 1200°F.                                  AND rising, THEN
                                                  ** GO TO 1-SACRG-1, Severe Accident Control Room Guideline Initial Response.
: 20. CHECK reactor subcritical:                CONTINUE to borate.
: a. Power range channels less than       IF boration is NOT available, 5%.                                  THEN ALLOW RCS to heat up to insert
: b. Intermediate range startup rate negative reactivity from temperature NEGATIVE.
coefficients.
IF red OR orange condition exists on other Status Trees, THEN PERFORM actions of other FR Procedures which do not cool down or otherwise add positive reactivity to the core.
                                                  ** GO TO Step 4.
: 21. TERMINATE emergency boration:
: a. PLACE BA transfer pumps in SLOW speed.
: b. CLOSE emergency borate valve 1-FCV-62-138.
: c. IF alternate boration opened, THEN Locally CLOSE 1-ISV-62-929.
Page 10 of 16
 
STEP DESCRIPTION TABLE FOR FR-S.l                  Step __2__
STEP:       Verify Turbine Trip This is a time critical step (within PURPOSE:   To ensure that the turbine is tripped      30 seconds).
BASIS:
The turbine is tripped to prevent an uncontrolled cool down of the RCS due to steam flow that the turbine would require. For an ATWS event where a loss of normal feedwater has occurred, analyses have shown that a turbine trip is necessary (within 30 seconds) to maintain SG inventory.
If the turbine will not trip, a turbine runback (manual decrease in load) at maximum rate will also reduce steam flow in a delayed manner. If the turbine stop valves cannot be closed by either trip or runback, the MSIVs should be closed.
ACTIONS:
o    Determine if all turbine stop valves are closed o    Determine if turbine will not trip o    Determine if turbine cannot be run back o    Trip the turbine o  Manually run back turbine o    Close main steamline isolation and bypass valves INSTRUMENTATION:
o    Turbine stop valve position indication o    MSIVs and bypass valves position indication CONTROL/EQUIPMENT:
o    Switches for turbine trip (e.g. manual trip buttons, overspeed test switch, EH control oil pump switches) o    Controls to manually run back turbine o    Switches to close MSIVs and bypass valves FR-S.l Background                    77                HP-Rev. 2, 4/30/2005 HFRSIBG.doc
 
STEP DESCRIPTION TABLE FOR FR-S.l              Step    3__
STEP:       Check AFW Pumps Running This is another time critical step PURPOSE:   To ensure AFW pumps are running                    (within 60 seconds).
BASIS:
The MD AFW pumps start automatically on an SI signal and SG low level to provide feed to the SGs for decay heat removal. If SG levels drop below the appropriate setpoint, the turbine-driven AFW pump will also automatically start to supplement the MD pumps. The ATWS analyses have shown that actuation of AFW within 60 seconds after the failure to scram provides acceptable results.
ACTIONS:
o  Determine if MD AFW pumps are running o  Determine if the turbine-driven AFW pump is running if necessary o  Start MD AFW pumps o  Open steam supply valves to turbine-driven AFW pump INSTRUMENTATION:
o  MD AFW pumps status indication o  Turbine-driven AFW pump status indication o  Turbine-driven AFW pump steam supply valve position indication CONTROL/EQUIPMENT:
Switches for:
o MD AFW pumps o Turbine-driven AFW pump steam supply valves KNOWLEDGE:
N/A PLANT-SPECIFIC INFORMATION:
N/A FR-S.l Background                    79                HP-Rev. 2, 4/30/2005 HFRSIBG.doc
 
Notice that even though this is "the most direct manner of adding negative STEP DESCRIPTION TABLE FOR FR-S.l            reactivity Step 4__  to the core," it is not a time critical step.
STEP:       Initiate Emergency Boration of RCS PURPOSE:   To add negative reactivity to bring the reactor core subcritical BASIS:
After control rod trip and rod insertion functions, boration is the next most direct manner of adding negative reactivity to the core. The intended boration path here is the most direct one available, not requiring SI initiation, but using normal charging pump(s). Pump miniflow lines are assumed to be open to protect the pumps in the event of high RCS pressure.
Several plant specific means are usually available for rapid boration and should be specified here in order of preference. Methods of rapid boration include emergency boration, injecting the BIT, and safety injection actuation.
It should be noted that SI actuation will trip the main feedwater pumps. If this is undesirable, the operator can manually align the system for safety injection. However, the RWST valves to the suction of the SI pumps should be opened first before opening up the BIT valves. If a safety injection is already in progress but is having no effect on nuclear flux, then the BIT and RWST are not performing their intended function, perhaps due to blockage or leakage. In this case some other alignment using the BATs and/or non-safeguards charging pump(s) is required.
The check on RCS pressure is intended to alert the operator to a condition which would reduce charging or SI pump injection into the RCS and, therefore, boration. The PRZR PORV pressure setpoint is chosen as that pressure at which flow into the RCS is insufficient. The contingent action is a rapid depressurization to a pressure which would allow increased injection flow.
When primary pressure drops 200 psi below the PORV pressure setpoint, the PORVs should be closed. The operator must verify successful closure of the PORVs, closing the isolation valves, if necessary.
FR-S.l Background                    80                HP-Rev. 2, 4/30/2005 HFRSIBG.doc
 
STEP DESCRIPTION TABLE FOR FR-S.l            Step  4__
ACTIONS:
o  Determine if PRZR pressure is less than (A.02) psig o  Determine if PRZR PORVs and block valves are open o  Start charging/SI pumps o  Start PO pump o  Align boration path o  Align charging flow path o  Open PRZR PORVs and block valves as necessary until PRZR pressure is less than (A.08) psig INSTRUMENTATION:
o  Charging/SI pump(s) status indication o    PO pump status indication o  Position indication for charging path valves, boration path valves o    PRZR pressure indication o    PRZR PORV and block valve position indications CONTROL/EQUIPMENT:
o  Charging/SI pump(s) switches o    PO pump switch o  Switches for charging path valves/boration path valves o    PRZR PORVs and block valves switches KNOWLEDGE:
N/A PLANT-SPECIFIC INFORMATION:
o   (A.02) PRZR PORV pressure setpoint.
o   (A.08) 200 psi less than PRZR PORV pressure setpoint.
o  Preferred alignments for emergency boration based on plant equipment and operating practices.
FR-S.1 Background                    81                HP-Rev. 2, 4/30/2005 HFRS1BG.doc
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]
Some examples of SRO exam items for this topic include:
* 10 CFR 50.59 screening and evaluation processes.
* 10 CFR 50.59 screening and evaluation processes.
* Administrative processes for temporary modifications.
* Administrative processes for temporary modifications.
* Administrative processes for disabling annunciators.
* Administrative processes for disabling annunciators.
* Administrative processes for the installation of temporary instrumentation.
* Administrative processes for the installation of temporary instrumentation.
* Processes for changing the plant or plant procedures. Section IV provides an example of a satisfactory SRO-only question related to this topic. D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.       [10 CFR 55.43(b)(4)] Some examples of SRO exam items for this topic include:
* Processes for changing the plant or plant procedures.
Section IV provides an example of a satisfactory SRO-only question related to this topic.
D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
[10 CFR 55.43(b)(4)]
Some examples of SRO exam items for this topic include:
* Process for gaseous/liquid release approvals, i.e., release permits.
* Process for gaseous/liquid release approvals, i.e., release permits.
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures. Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits. SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc. E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedure's overall mitigative strategy or purpose.  
Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.
 
SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.
The applicant's knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedure's content is required to correctly answer the written test item, for example:
E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.
The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:
Page 6 of 16


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 7 of 16
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
* Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
* Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
SRO-only knowledge should not be claimed for questions that can be answered solely using "systems knowledge"; e.g.:
SRO-only knowledge should not be claimed for questions that can be answered solely using systems knowledge; e.g.:
* how the system works.
* how the system works.
* system flow path.
* system flow path.
* component locations, etc. SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:
* component locations, etc.
SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:
* the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
* the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
* any AOP entry condition.
* any AOP entry condition.
* plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
* plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
* immediate operator actions of a procedure. Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.        
* immediate operator actions of a procedure.
Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.
Page 7 of 16
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)
(Assessment and selection of procedures)
Can the question be answered solely by knowing systems knowledge, i.e., how the system works,      Yes RO question flowpath, logic, component location?
No Can the question be answered solely by knowing immediate operator actions?                            Yes    RO question No Can the question be answered solely by knowing entry conditions for AOPs or plant parameters          Yes RO question that require direct entry to major EOPs?
No Can the question be answered solely by knowing the purpose, overall sequence of events, or            Yes RO question overall mitigative strategy of a procedure?
No Does the question require one or more of the following?
* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps                                            Yes        SRO-only
* Knowledge of diagnostic steps and decision points in the              question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 8 of 16  Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5) (Assessment and selection of procedures)       
78.
Given the following conditions:
    -  The Unit 1 #1 SG is ruptured.
    -  The RCPs are SECURED.
    -  In accordance with step 18 of 1-E-3, the crew has INITIATED RCS cooldown to the target incore temperature.
    -  An ORANGE path exists for 1-FR-P.1 based SOLELY upon the Tcold in Loop #1.
Which ONE of the following describes the procedure transition requirements in accordance with 1-E-3?
The Unit Supervisor WILL __________.
NOTE: 1-E-3, Steam Generator Tube Rupture 1-FR-P.1, Pressurized Thermal Shock A. IMMEDIATELY transition to 1-FR-P.1 B. NOT transition to 1-FR-P.1 UNTIL 1-E-3 is completed C. transition to 1-FR-P.1 IF the ORANGE path still exists ONCE the cooldown to target incore temperature is completed D. transition to 1-FR-P.1 IF the ORANGE path still exists ONCE SI is terminated in accordance with 1-E-3


Can the question be answered solely by knowing "systems knowledge", i.e., how the system works, flowpath, logic, component location? RO question YesNoCan the question be answered solely by knowing immediate operator actions? YesCan the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?YesNoDoes the question require one or more of the following?
CORRECT ANSWER:                                                               D DISTRACTOR ANALYSIS:
* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
A. Incorrect: As seen in 1-E-3, 1-E-3, Steam Generator Tube Rupture, If RCPs are NOT running, a false red or orange path may be indicated for 1-FR-P.1 during the following steps. T-cold in the ruptured loop should be disregarded until Step 43. Steps 32 to 42 of 1-E-3, stop the safety injection, realign normal charging and letdown and restore normal pressure control. Therefore, a transition to 1-FR-P.1 is not allowed until the SI is terminated in accordance with 1-E-3. It is plausible to believe that a transition to 1-FR-P.1 would be immediately effected as if either a red or orange path is indicated on a status tree, then a transition to that trees restoration procedure is normally mandated.
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps
B. Incorrect: Again, a transition to 1-FR-P.1 is not allowed until after safety injection is terminated in accordance with 1-E-3. It is plausible to believe that a transition would be delayed until after 1-E-3 is completed because one may recall that a restriction on the use of 1-FR-P.1 exists and then misapply such.
* Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
C. Incorrect: This distractor is also incorrect and plausible for the same reason as the B distractor.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?NoNoYesQuestion might not be linked to 10 CFR 55.43(b)(5) for SRO-only NoRO question RO question RO question 
D. Correct: As described, it is correct that the use of 1-FR-P.1 is not allowed until after SI is terminated in accordance with 1-E-3.
: 78. Given the following conditions: - The Unit 1 #1 SG is ruptured. - The RCPs are SECURED. - In accordance with step 18 of 1-E-3, the crew has INITIATED RCS cooldown to the target incore temperature.
- An ORANGE path exists for 1-FR-P.1 based SOLELY upon the Tcold in Loop #1. Which ONE of the following describes the procedure transition requirements in accordance with 1-E-3? The Unit Supervisor WILL __________. NOTE: 1-E-3, Steam Generator Tube Rupture  1-FR-P.1, Pressurized Thermal Shock  A. IMMEDIATELY transition to 1-FR-P.1 B. NOT transition to 1-FR-P.1 UNTIL 1-E-3 is completed C. transition to 1-FR-P.1 IF the ORANGE path still exists ONCE the cooldown to target incore temperature is completed D. transition to 1-FR-P.1 IF the ORANGE path still exists ONCE SI is terminated in accordance with 1-E-3 CORRECT ANSWER:D DISTRACTOR ANALYSIS:   A. Incorrect: As seen in 1-E-3, 1-E-3, Steam Generator Tube Rupture, "If RCPs are NOT running, a false red or orange path may be indicated for 1-FR-P.1 during the following steps. T-cold in the ruptured loop should be disregarded until Step 43."  Steps 32 to 42 of 1-E-3, stop the safety injection, realign normal charging and letdown and restore normal pressure control. Therefore, a transition to 1-FR-P.1 is not allowed until the SI is terminated in accordance with 1-E-3. It is plausible to believe that a transition to 1-FR-P.1 would be immediately effected as if either a red or orange path is indicated on a status tree, then a transition to that tree's restoration procedure is normally mandated. B. Incorrect: Again, a transition to 1-FR-P.1 is not allowed until after safety injection is terminated in accordance with 1-E-3. It is plausible to believe that a transition would be delayed until after 1-E-3 is completed because one may recall that a restriction on the use of 1-FR-P.1 exists and then misapply such. C. Incorrect: This distractor is also incorrect and plausible for the same reason as the B distractor. D. Correct: As described, it is correct that the use of 1-FR-P.1 is not allowed until after SI is terminated in accordance with 1-E-3.
Question Number: 78 Tier:  1 Group:  1 K/A: 038 Steam Generator Tube Rupture 2.4 Emergency Procedures / Plan 2.4.31 Knowledge of annunciator alarms, indications, or response procedures.
Importance Rating: 4.2  4.1  10 CFR Part 55: (CFR: 41.10 / 43.5 / 45.11)


10CFR55.43.b: 10 CFR 55.43(b)(5) K/A Match: K/A is matched because the applicant is required to correctly implement the procedures 1-E-3 and 1-FR-P.1 during a SGTR. Technical  
Question Number:      78 Tier:    1  Group:      1 K/A:    038 Steam Generator Tube Rupture 2.4 Emergency Procedures / Plan 2.4.31 Knowledge of annunciator alarms, indications, or response procedures.
Importance Rating:    4.2 4.1 10 CFR Part 55:      (CFR: 41.10 / 43.5 / 45.11) 10CFR55.43.b:       10 CFR 55.43(b)(5)
K/A Match:   K/A is matched because the applicant is required to correctly implement the procedures 1-E-3 and 1-FR-P.1 during a SGTR.
1-E-3, Steam Generator Tube Rupture Technical  


==Reference:==
==Reference:==
1-E-3, Steam Generator Tube Rupture TI-12.04, User's Guide for Abnormal and Emergency Operating Instructions Proposed references to be provided: None  Learning Objective: 3-OT-TI1204 24. State the action required when a RED or Orange Path is diagnosed while monitoring the CSF status trees. Cognitive Level:     Higher X Lower     Question Source:     New X Modified Bank   Bank     Question History: New question for the 2015-301 NRC SRO Exam Comments: See the marked up Clarification Guidance for SRO-only Questions.  
TI-12.04, Users Guide for Abnormal and Emergency Operating Instructions Proposed references to   None be provided:
Learning Objective:       3-OT-TI1204
: 24. State the action required when a RED or Orange Path is diagnosed while monitoring the CSF status trees.
Cognitive Level:
Higher             X Lower Question Source:
New               X Modified Bank Bank Question History:         New question for the 2015-301 NRC SRO Exam Comments:                 See the marked up Clarification Guidance for SRO-only Questions.


WBN Unit 1 Steam Generator Tube Rupture 1-E-3 Rev. 0003   Step    Action/Expected Response   Response Not Obtained   Page 11 of 47    CAUTION
WBN             Steam Generator Tube Rupture             1-E-3 Unit 1                                                    Rev. 0003 Step    Action/Expected Response                   Response Not Obtained This is the caution immediately prior to the steps directing the initial cooldown to target incore temp.
CAUTION
* The 1500 psig RCP trip criteria is NOT applicable during or after a controlled RCS cooldown and depressurization.
* The 1500 psig RCP trip criteria is NOT applicable during or after a controlled RCS cooldown and depressurization.
* If total feed flow CAPABILITY of 410 gpm is AVAILABLE, 1-FR-H.1, Loss of Secondary Heat Sink, should NOT be implemented.
* If total feed flow CAPABILITY of 410 gpm is AVAILABLE, 1-FR-H.1, Loss of Secondary Heat Sink, should NOT be implemented.
* Excessive steam dump cooldown rate will cause MSIV isolation due to the rate sensitive signal.
* Excessive steam dump cooldown rate will cause MSIV isolation due to the rate sensitive signal.
* If RCPs are NOT running, a false red or orange path may be indicated for 1-FR-P.1 during the following steps. T-cold in the ruptured loop should be disregarded until Step 43.
* If RCPs are NOT running, a false red or orange path may be indicated for 1-FR-P.1 during the following steps. T-cold in the ruptured loop should be disregarded until Step 43.
WBN Unit 1 Steam Generator Tube Rupture 1-E-3 Rev. 0003  Step    Action/Expected Response   Response Not Obtained   Page 12 of 47  18. INITIATE RCS cooldown to target Incore temp, determined from Step 17. a. DUMP steam to condenser from Intact S/G(s) at maximum achievable rate:  
If RCPs are not running, do not transition to 1-FR-P.1 until AFTER SI is terminated.
Page 11 of 47
 
WBN               Steam Generator Tube Rupture             1-E-3 Unit 1                                                      Rev. 0003 Step   Action/Expected Response                   Response Not Obtained START THE
: 18. INITIATE RCS cooldown to                                             INITIAL target Incore temp, determined                                       COOLDOWN from Step 17.
: a. DUMP steam to condenser from           a. IF condenser steam dumps NOT Intact S/G(s) at maximum                   available, THEN achievable rate:
USE Intact S/G PORVs at IF dumps are in Tavg mode,                  maximum achievable cooldown THEN:                                      rate.
: 1)  PLACE steam dump controls OFF.                          IF an Intact S/G is NOT available,
: 2)  PLACE steam dump mode                  THEN switch in STEAM PRESSURE.                              PERFORM one BUT NOT BOTH of the following:
: 3)  ENSURE steam dump demand indicator 1-XI-1-33
* USE Faulted S/G, reading zero.
: 4)  PLACE steam dump                            OR controls ON.                          *    ** GO TO 1-ECA-3.1, SGTR
: 5)  PLACE steam dump                            LOCA - Subcooled Recovery.
controller in MAN, AND FULLY OPEN three cooldown valves
( 25% demand).
Step continued on the next page Page 12 of 47


IF dumps are in Tavg mode, THEN: 1) PLACE steam dump controls OFF. 2) PLACE steam dump mode switch in STEAM PRESSURE. 3) ENSURE steam dump demand indicator 1-XI-1-33 reading zero. 4) PLACE steam dump controls ON. 5) PLACE steam dump controller in MAN, AND  FULLY OPEN three cooldown valves  ( 25% demand). a. IF condenser steam dumps NOT available, THEN  USE Intact S/G PORVs at maximum achievable cooldown rate. IF an Intact S/G is NOT available, THEN  PERFORM one BUT NOT BOTH of the following:
WBN               Steam Generator Tube Rupture             1-E-3 Unit 1                                                    Rev. 0003 Step   Action/Expected Response                     Response Not Obtained CAUTION
* USE Faulted S/G,  OR * **  GO TO 1-ECA-3.1, SGTR LOCA - Subcooled Recovery. Step continued on the next page WBN Unit 1 Steam Generator Tube Rupture 1-E-3 Rev. 0003  Step    Action/Expected Response   Response Not Obtained   Page 21 of 47    CAUTION
* SI should be terminated as quickly as possible after termination criteria are met to prevent Ruptured S/G overfill.
* SI should be terminated as quickly as possible after termination criteria are met to prevent Ruptured S/G overfill.
* If total feed flow CAPABILITY of 410 gpm is AVAILABLE, 1-FR-H.1, Loss of Secondary Heat Sink, should NOT be implemented. 32. CHECK SI termination criteria: a. CHECK RCS subcooling greater than 65°F [85°F ADV]. a. ** GO TO 1-ECA-3.1, SGTR and LOCA - Subcooled Recovery. b. CHECK secondary heat sink  with either:
* If total feed flow CAPABILITY of 410 gpm is AVAILABLE, 1-FR-H.1, Loss of Secondary Heat Sink, should NOT be implemented.
* Total available feed flow greater than 410 gpm,   OR
TERMINATE THE SI.
* At least one S/G NR level greater than 29%
: 32. CHECK SI termination criteria:
[39% ADV]. b. **  GO TO 1-FR-H.1, Loss of  Heat Sink. c. CHECK RCS pressure stable or rising. c. ** GO TO 1-ECA-3.1, SGTR and LOCA - Subcooled Recovery. d. CHECK PZR level greater  than 15% [33% ADV]. d. **  GO TO Step 16.
: a. CHECK RCS subcooling greater             a. ** GO TO 1-ECA-3.1, SGTR and than 65°F [85°F ADV].                        LOCA - Subcooled Recovery.
WBN Unit 1 Steam Generator Tube Rupture 1-E-3 Rev. 0003   Step   Action/Expected Response   Response Not Obtained   Page 28 of 47  42. (continued) d. MAINTAIN RCS pressure at Ruptured S/G pressure:
: b. CHECK secondary heat sink               b. ** GO TO 1-FR-H.1, Loss of with either:                                 Heat Sink.
* CONTROL PZR heaters  as necessary.
* Total available feed flow greater than 410 gpm, OR
* USE normal PZR spray  as necessary. d. IF letdown in service, THEN ALIGN aux spray USING Appendix A (1-E-3) ALIGN AUX SPRAY. IF letdown NOT in service, THEN USE one PZR PORV, AND MONITOR the following:
* At least one S/G NR level greater than 29%
[39% ADV].
: c. CHECK RCS pressure                       c. ** GO TO 1-ECA-3.1, SGTR and stable or rising.                            LOCA - Subcooled Recovery.
: d. CHECK PZR level greater                 d. ** GO TO Step 16.
than 15% [33% ADV].
STEPS 33 to 42:
stop the safety injection realign normal charging and letdown regain pressure control Page 21 of 47
 
WBN               Steam Generator Tube Rupture             1-E-3 Unit 1                                                      Rev. 0003 Step     Action/Expected Response                   Response Not Obtained
: 42.     (continued)
: d. MAINTAIN RCS pressure at             d. IF letdown in service, THEN Ruptured S/G pressure:
ALIGN aux spray USING
* CONTROL PZR heaters Appendix A (1-E-3) as necessary.
ALIGN AUX SPRAY.
* USE normal PZR spray as necessary.
IF letdown NOT in service, THEN This is the first time (since just before                                      USE one PZR PORV, AND the start of the initial cooldown to                                      MONITOR the following:
target incore
* Vessel head void formation.
* Vessel head void formation.
temperature) that a transition to 1-FR-
* PZR level rise.
* PZR level rise.
* PRT rupture. NOTE Normal monitoring of T-cold for 1-FR-P.1 can now be resumed. The Caution prior to Step 18 regarding a false red or orange path is no longer applicable. 43. DETERMINE if Cntmt spray should be stopped: a. MONITOR Cntmt pressure less than 2.0 psig. a. WHEN Cntmt pressure  less than 2.0 psig, THEN PERFORM Substeps 43b thru e.
P.1 would be
** GO TO Step 44. b. CHECK at least one Cntmt spray pump running. b. ** GO TO Step 44. Step continued on next page Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 6 of 16  C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)] Some examples of SRO exam items for this topic include:
* PRT rupture.
permitted.
NOTE           Normal monitoring of T-cold for 1-FR-P.1 can now be resumed.
The Caution prior to Step 18 regarding a false red or orange path is no longer applicable.
: 43.         DETERMINE if Cntmt spray should be stopped:
: a. MONITOR Cntmt pressure               a. WHEN Cntmt pressure less than 2.0 psig.                         less than 2.0 psig, THEN PERFORM Substeps 43b thru e.
                                                            ** GO TO Step 44.
: b. CHECK at least one                   b.   ** GO TO Step 44.
Cntmt spray pump running.
Step continued on next page Page 28 of 47
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]
Some examples of SRO exam items for this topic include:
* 10 CFR 50.59 screening and evaluation processes.
* 10 CFR 50.59 screening and evaluation processes.
* Administrative processes for temporary modifications.
* Administrative processes for temporary modifications.
* Administrative processes for disabling annunciators.
* Administrative processes for disabling annunciators.
* Administrative processes for the installation of temporary instrumentation.
* Administrative processes for the installation of temporary instrumentation.
* Processes for changing the plant or plant procedures. Section IV provides an example of a satisfactory SRO-only question related to this topic. D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.       [10 CFR 55.43(b)(4)] Some examples of SRO exam items for this topic include:
* Processes for changing the plant or plant procedures.
Section IV provides an example of a satisfactory SRO-only question related to this topic.
D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
[10 CFR 55.43(b)(4)]
Some examples of SRO exam items for this topic include:
* Process for gaseous/liquid release approvals, i.e., release permits.
* Process for gaseous/liquid release approvals, i.e., release permits.
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures. Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits. SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc. E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedure's overall mitigative strategy or purpose.  
Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.
 
SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.
The applicant's knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedure's content is required to correctly answer the written test item, for example:
E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
The question requires the        This 10 CFR 55.43 topic involves both 1) assessing plant conditions detailed knowledge  (normal, abnormal, or emergency) and then 2) selecting a procedure or of a note in the    section of a procedure to mitigate, recover, or with which to proceed. One procedure versus    area of SRO level knowledge (with respect to selecting a procedure) is the overall          knowledge of the content of the procedure versus knowledge of the strategy.            procedures overall mitigative strategy or purpose.
The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:
Page 6 of 16


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 7 of 16
The question requires the                Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) knowledge of the decision point of
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
* Knowledge of when to implement attachments and appendices, including whether or not to      how to coordinate these items with procedure steps.
* Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
transition to 1-FR-
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
* Knowledge of diagnostic steps and decision points in the emergency P.1 (e.g. the          operating procedures (EOP) that involve transitions to event specific sub-detailed knowledge      procedures or emergency contingency procedures.
SRO-only knowledge should not be claimed for questions that can be answered solely using "systems knowledge"; e.g.:
of the point in the
* Knowledge of administrative procedures that specify hierarchy, procedure).            implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
SRO-only knowledge should not be claimed for questions that can be answered solely using systems knowledge; e.g.:
* how the system works.
* how the system works.
* system flow path.
* system flow path.
* component locations, etc. SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:
Note that the
* the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
* component locations, etc.
question does not require the          SRO-only knowledge should not be claimed for questions that can be fundamental          answered solely using fundamental knowledge of:
knowledge of the plant parameters
* the basic purpose, the overall sequence of events that will occur, or the requiring entry into    overall mitigative strategy of a procedure.
1-FR-P.1 (i.e.
* any AOP entry condition.
* any AOP entry condition.
* plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
requiring the RO
* immediate operator actions of a procedure. Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.        
* plant parameters that require direct entry to major EOPs; e.g., major LOK).                    Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
* immediate operator actions of a procedure.
Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.
Page 7 of 16
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)
(Assessment and selection of procedures)
Can the question be answered solely by knowing systems knowledge, i.e., how the system works,      Yes RO question flowpath, logic, component location?
No Can the question be answered solely by knowing immediate operator actions?                            Yes    RO question No Can the question be answered solely by knowing entry conditions for AOPs or plant parameters          Yes RO question that require direct entry to major EOPs?
No Can the question be answered solely by knowing the purpose, overall sequence of events, or            Yes RO question overall mitigative strategy of a procedure?
No Does the question require one or more of the following?
* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps                                            Yes        SRO-only
* Knowledge of diagnostic steps and decision points in the              question This bullet could EOPs that involve transitions to event specific sub-also be utilized as procedures or emergency contingency procedures TI-12.04 also
* Knowledge of administrative procedures that specify delineates the use      hierarchy, implementation, and/or coordination of plant of 1-FR-P.1 during      normal, abnormal, and emergency procedures                      Again, the 1-E-3.                                                                                  Question requires No                                                      the detailed knowledge of a note (e.g. a Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only                          decision point -
whether or not to implement 1-FR-P.1).
Page 8 of 16


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 8 of 16  Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5) (Assessment and selection of procedures)        
79.
Given the following timeline:
00:00:00 0-SI-0-3, Weekly Log has been completed and indicates the following TWO items:
ITEM ONE ITEM TWO 00:00:0 1 1-XS-57-96, 125 VITAL BATT BD VOLTMETER SELECTOR is in position I 00: 01:0 1 1-EI-57-96, VIT BATT BDS VOLTS reads 131.5 VDC 00: 1 1 :00 1-EI-57-96 reads 127.5 VDC 00:21:00 1-EI-57-96 reads 123.5 VDC Which ONE of the following describes the FIRST time that LCO 3.8.4, DC Sources -
Operating will NOT be met??
T/S LCO 3.8.4 will FIRST NOT met at time ________.
A.      00:00:00 B.      00:0 1:00 C.       00: 1 1 :00 D.        00:2 1:00


Can the question be answered solely by knowing "systems knowledge", i.e., how the system works, flowpath, logic, component location? RO question YesNoCan the question be answered solely by knowing immediate operator actions? YesCan the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?YesNoDoes the question require one or more of the following?
CORRECT ANSWER:                                                                A DISTRACTOR ANALYSIS:
* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
A. Correct: As given in the stem of the question, at 00:00:00, data recorded on 0-SI-03 reveals that the CB 2 (the output breaker for 0-CHGR-236-1) is A. This is the nomenclature which stipulates that it is available but not closed or inoperable. Also, breaker 225 (the 125V Vital Batt Bd I breaker which ties a spare battery charger to the board) is available.
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps
Because of these facts, one may ascertain that NO charger is aligned to the Vital Battery Board. The LCO bases for T/S LCO 3.8.4 stipulate, An OPERABLE vital DC electrical power subsystem requires all required batteries and respective chargers to be operating and connected to the associated DC buses. Therefore, because NO charger is aligned to the Vital Batt Bd I and such information was received by the SRO at 00:00:00, actions of T/S LCO 3.8.4 are required at 00:00:00. Additionally, actions of T/S LCO 3.8.4 are necessary through the declaration of SR 3.0.1 which stipulates, Failure to meet the surveillanceshall be failure to meet the LCO.
* Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
B. Incorrect: This distractor is incorrect because as mentioned the required actions of the T/S LCO were first required at 00:00:00. It is plausible however because if one did not understand the content of the basis for T/S LCO 3.8.4 and subsequently applied the voltage criteria for Vital Battery V to Vital Battery I, one would arrive at the result that this distractor was correct.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?NoNoYesQuestion might not be linked to 10 CFR 55.43(b)(5) for SRO-only NoRO question RO question RO question 
C. Incorrect: This distractor is incorrect because as mentioned the required actions of the T/S LCO were first required at 00:00:00. It is plausible however because if one did not understand the content of the basis for T/S LCO 3.8.4 and subsequently applied the correct voltage criteria for Vital Battery I, one would arrive at the result that this distractor was correct.
: 79. Given the following timeline: 00:00:000-SI-0-3, Weekly Log has been completed and indicates the following TWO items: ITEM ONE  ITEM TWO  00:00:011-XS-57-96, 125 VITAL BATT BD VOLTMETER SELECTOR is in position I 00:01:011-EI-57-96, VIT BATT BDS VOLTS reads 131.5 VDC 00:11:001-EI-57-96 reads 127.5 VDC 00:21:001-EI-57-96 reads 123.5 VDC  Which ONE of the following describes the FIRST time that LCO 3.8.4, DC Sources - Operating will NOT be met??
D. Incorrect: This distractor is incorrect because as mentioned the required actions of the T/S LCO were first required at 00:00:00. It is plausible however because if one did not understand the content of the basis for T/S LCO 3.8.4 and subsequently applied the voltage criteria for the DG battery to Vital Battery I, one would arrive at the result that this distractor was correct.


T/S LCO 3.8.4 will FIRST NOT met at time ________. A. 00:00:00B. 00:0 1:00 C. 00:11:00D. 00:2 1:00 CORRECT ANSWER:A DISTRACTOR ANALYSIS:  A. Correct: As given in the stem of the question, at 00:00:00, data recorded on 0-SI-03 reveals that the CB 2 (the output breaker for 0-CHGR-236-1) is A. This is the nomenclature which stipulates that it is available but not closed or inoperable. Also, breaker 225 (the 125V Vital Batt Bd I breaker which ties a spare battery charger to the board) is available.
Question Number:       79 Tier:     1   Group:       1 K/A:   058 Loss of DC Power 2.4 Emergency Procedures / Plan 2.4.47 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.
Because of these facts, one may ascertain that NO charger is aligned to the Vital Battery Board. The LCO bases for T/S LCO 3.8.4 stipulate, An OPERABLE vital DC electrical power subsystem requires all required batteries and respective chargers to be operating and connected to the associated DC buses. Therefore, because NO charger is aligned to the Vital Batt Bd I and such information was received by the SRO at 00:00:00, actions of T/S LCO 3.8.4 are required at 00:00:00. Additionally, actions of T/S LCO 3.8.4 are necessary through the declaration of SR 3.0.1 which stipulates, Failure to meet the surveillanceshall be failure to meet the LCO. B. Incorrect: This distractor is incorrect because as mentioned the required actions of the T/S LCO were first required at 00:00:00. It is plausible however because if one did not understand the content of the basis for T/S LCO 3.8.4 and subsequently applied the voltage criteria for Vital Battery V to Vital Battery I, one would arrive at the result that this distractor was correct. C. Incorrect: This distractor is incorrect because as mentioned the required actions of the T/S LCO were first required at 00:00:00. It is plausible however because if one did not understand the content of the basis for T/S LCO 3.8.4 and subsequently applied the correct voltage criteria for Vital Battery I, one would arrive at the result that this distractor was correct. D. Incorrect: This distractor is incorrect because as mentioned the required actions of the T/S LCO were first required at 00:00:00. It is plausible however because if one did not understand the content of the basis for T/S LCO 3.8.4 and subsequently applied the voltage criteria for the DG battery to Vital Battery I, one would arrive at the result that this distractor was correct.
Importance Rating:     4.2 4.2 10 CFR Part 55:       (CFR: 41.10 / 43.5 / 45.12) 10CFR55.43.b:         10 CFR 55.43(b)(2)
Question Number: 79 Tier:  1 Group:   1 K/A: 058 Loss of DC Power 2.4 Emergency Procedures / Plan 2.4.47 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.
K/A Match:   K/A is matched because while in a loss of DC power the applicant is required to accurately diagnose the operability of the Vital Battery I.
Importance Rating: 4.2 4.2 10 CFR Part 55: (CFR: 41.10 / 43.5 / 45.12) 10CFR55.43.b: 10 CFR 55.43(b)(2) K/A Match: K/A is matched because while in a loss of DC power the applicant is required to accurately diagnose the operability of the Vital Battery I.
The applicant must do so using 0-SI-0-3, T/S LCO 3.8.4 and the data trended in a timeline contained in the stem of the question. The question is applicable to the loss of DC power because the loss of a vital charger is an initiator to such casualty.
The applicant must do so using 0-SI-0-3, T/S LCO 3.8.4 and the data trended in a timeline contained in the stem of the question. The question is applicable to the loss of DC power because the loss of a vital charger is an initiator to such casualty.
Technical  
Technical  


==Reference:==
==Reference:==
T/S LCO 3.8.4, DC Sources - Operating T/S LCO 3.8.4 Basis   Proposed references to be provided: None  Learning Objective:   Cognitive Level:     Higher X Lower     Question Source:     New X Modified Bank   Bank     Question History: New question for the 2015-301 NRC SRO Exam Comments: See the marked up Clarification Guidance for SRO-only Questions.
T/S LCO 3.8.4, DC Sources - Operating T/S LCO 3.8.4 Basis Proposed references to       None be provided:
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 3 of 16  II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]: A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)] Some examples of SRO exam items for this topic include:
Learning Objective:
Cognitive Level:
Higher               X Lower Question Source:
New                   X Modified Bank Bank Question History:           New question for the 2015-301 NRC SRO Exam Comments:                   See the marked up Clarification Guidance for SRO-only Questions.
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]:
A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)]
Some examples of SRO exam items for this topic include:
* Reporting requirements when the maximum licensed thermal power output is exceeded.
* Reporting requirements when the maximum licensed thermal power output is exceeded.
* Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems, fire doors, etc.
* Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems, fire doors, etc.
* The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
* The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
* National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
* National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
* Processes for TS and FSAR changes. Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic. B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)] Some examples of SRO exam items for this topic include:
* Processes for TS and FSAR changes.
* Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic.
* Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]
* Knowledge of TS bases that are required to analyze TS required actions and terminology.
Some examples of SRO exam items for this topic include:
The fact that a charger must be
* Application of Required Actions (Section 3) and Surveillance aligned for the DC          Requirements (SR) (Section 4) in accordance with rules of application source to be                requirements (Section 1).
considered
* Application of generic Limiting Condition for Operation (LCO)
OPERABLE is                requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
contained in the TS
* Knowledge g of TS bases that are required to analyze TS required actions basis.                      and terminology.
* Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).
* Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of1 hour action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.  
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of  1 hour action statements and the All of the            safety limits since Reactor Operators (ROs) are typically required to know distractors rely on    these items.
information contained "below      SRO-only knowledge generally cannot be claimed for questions that can be the line."            answered solely based on expected RO TS knowledge. ROs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Page 3 of 16


SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. ROs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the B. Facility operating limitations in the TS and their bases.[10 CFR 55.43(b)(2)] Knowledge of TS bases that are required to analyze TS required actions gand terminology. Thefactthatachargermustbe alignedfortheDC sourcetobe considered OPERABLEis containedintheTS basis.Allofthedistractorsrelyon information contained"below theline."
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 4 of 16 ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)         Above this line RO knowledge NotethatTS3.8.4onlyspecifiesthat DCsourcesmust beoperable.One mustlookinthe basistodetermine whataDCsource is.
ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 5 of 16 Figure 1:  Screening for SRO-only linked to 10 CFR 55.43(b)(2) (Tech Specs)       
Note that TS 3.8.4 only specifies that DC sources must be operable. One RO must look in the knowledge                                                          basis to determine what a DC source is.
Above this line Page 4 of 16


Can question be answered solely by knowing hour TS/TRM Action? RO question YesNoCan question be answered solely by knowing the LCO/TRM information listed above-the-line? YesRO question NoCan question be answered solely by knowing the TS Safety Limits? YesRO question NoDoes the question involve one or more of the following for TS, TRM, or ODCM?
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)
(Tech Specs)
Can question be answered solely by knowing  1      Yes hour TS/TRM Action?                                           RO question No Can question be answered solely by knowing the       Yes LCO/TRM information listed above-the-line?                   RO question No Can question be answered solely by knowing the       Yes TS Safety Limits?                                             RO question No Does the question involve one or more of the following for TS, TRM, or ODCM?
* Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
* Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)
* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4))                            Yes      SRO-only
* Knowledge of TS bases that is required to analyze TS required actions and terminology SRO-only question YesNoQuestion might not be linked to 10 CFR 55.43(b)(2) for SRO-only NoNoNo)Knowledge of TS bases that is required to analyze TSgrequired actions and terminology YesSRO-only question DC Sources - Operating 3.8.4        Watts Bar-Unit 1 3.8-24      3.8  ELECTRICAL POWER SYSTEMS 3.8.4  DC Sources - Operating
* Knowledge g of TS bases that is required to analyze TS                 question required actions and terminology No Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only Page 5 of 16


LCO 3.8.4 Four channels of vital DC and four Diesel Generator (DG) DC electrical power subsystems shall be OPERABLE.  
DC Sources - Operating 3.8.4 3.8 ELECTRICAL POWER SYSTEMS 3.8.4 DC Sources - Operating LCO 3.8.4               Four channels of vital DC and four Diesel Generator (DG) DC electrical power subsystems shall be OPERABLE.
No mention of the charger is made in the "above the          ------------------------------------------------NOTES------------------------------------------------
line" portion of the T/S.          1.        Vital Battery V may be substituted for any of the required vital batteries.
: 2.        The C-S DG and its associated DC electrical power subsystem may be substituted for any of the required DGs and their associated DC electrical power subsystem.
APPLICABILITY:          MODES 1, 2, 3, and 4.
ACTIONS CONDITION                                  REQUIRED ACTION                            COMPLETION TIME A.      One vital DC electrical            A.1        Restore vital DC electrical                  2 hours power subsystem                                power subsystem to inoperable.                                    OPERABLE status.
B.      Required Action and                B.1        Be in MODE 3.                                6 hours Associated Completion Time of Condition A not            AND met.
B.2        Be in MODE 5.                                36 hours C.      One DG DC electrical                C.1        Restore DG DC electrical                    2 hours power subsystem                                power subsystem to inoperable.                                    OPERABLE status.
(continued)
Watts Bar-Unit 1                                        3.8-24


------------------------------------------------NOTES------------------------------------------------
DC Sources - Operating 3.8.4 ACTIONS (continued)
: 1. Vital Battery V may be substituted for any of the required vital batteries. 
CONDITION                               REQUIRED ACTION                 COMPLETION TIME D.     Required Action and              D.1     Declare associated DG            Immediately associated Completion                    inoperable.
: 2. The C-S DG and its associated DC electrical power subsystem may be substituted for any of the required DGs and their associated DC electrical power subsystem. -----------------------------------------------------------------------------------------------------------  APPLICABILITY: MODES 1, 2, 3, and 4.
Time of Condition C not met.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One vital DC electrical power subsystem inoperable.
The S/Rs are the source of the voltage requirements.
A.1 Restore vital DC electrical power subsystem to OPERABLE status.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                            FREQUENCY SR 3.8.4.1             Verify vital battery terminal voltage is    128 V (132 V    7 days for vital battery V) on float charge.
2 hours B. Required Action and Associated Completion Time of Condition A not met. B.1 Be in MODE 3. AND B.2 Be in MODE 5. 6 hours   36 hours  C. One DG DC electrical power subsystem inoperable.
SR 3.8.4.2              Verify DG battery terminal voltage is      124 V on float 7 days charge.
C.1 Restore DG DC electrical power subsystem to OPERABLE status.
SR 3.8.4.3              Verify for the vital batteries that the alternate feeder   7 days breakers to each required battery charger are open.
2 hours  (continued)
SR 3.8.4.4              Verify correct breaker alignment and indicated power        7 days availability for each DG 125 V DC distribution panel and associated battery charger.
Nomentionofthechargerismadeinthe"abovethe line"portionoftheT/S.
(continued)
DC Sources - Operating 3.8.4        Watts Bar-Unit 1 3.8-25     ACTIONS  (continued) CONDITION REQUIRED ACTION COMPLETION TIME  D. Required Action and associated Completion Time of Condition C not met.
Watts Bar-Unit 1                                     3.8-25
D.1 Declare associated DG inoperable.
Immediately 


SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR  3.8.4.1  Verify vital battery terminal voltage is  128 V (132 V for vital battery V) on float charge. 7 days  SR  3.8.4.2  Verify DG battery terminal voltage is  124 V on float charge. 7 days  SR 3.8.4.3  Verify for the vital batteries that the alternate feeder breakers to each required battery charger are open. 7 days  SR  3.8.4.4  Verify correct breaker alignment and indicated power availability for each DG 125 V DC distribution panel and associated battery charger. 7 days  (continued)
DC Sources-Operating B 3.8.4 BASES APPLICABLE       The OPERABILITY of the DC sources is consistent with the initial assumptions SAFETY ANALYSES of the accident analyses and is based upon meeting the design basis of the (continued)     plant. This includes maintaining the DC sources OPERABLE during accident conditions in the event of:
TheS/Rsarethesourceofthevoltagerequirements.
: a.       An assumed loss of all offsite AC power or all onsite AC power; and
DC Sources-Operating B 3.8.4 BASES   (continued)    Watts Bar-Unit 1 B 3.8-57 Revision 113    APPLICABLE The OPERABILITY of the DC sources is consistent with the initial assumptions SAFETY ANALYSES of the accident analyses and is based upon meeting the design basis of the   (continued) plant. This includes maintaining the DC sources OPERABLE during accident conditions in the event of:
: b.       A worst case single failure.
: a. An assumed loss of all offsite AC power or all onsite AC power; and
The DC sources satisfy Criterion 3 of the NRC Policy Statement.
: b. A worst case single failure.  
LCO              Four 125V vital DC electrical power subsystems, each vital subsystem channel consisting of a battery bank, associated battery charger and the corresponding control equipment and interconnecting cabling supplying power to the associated Note that a charger          DC bus within the channel; and four DG DC electrical power subsystems each must be operating            consisting of a battery, a dual battery charger assembly, and the corresponding control equipment and interconnecting cabling are required to be OPERABLE to and connected.
ensure the availability of the required power to shut down the reactor and maintain it in a safe condition after an anticipated operational occurrence (A00) or a postulated DBA. Loss of any DC electrical power subsystem does not prevent the minimum safety function from being performed (Ref. 4).
An OPERABLE vital DC electrical power subsystem requires all required batteries and respective chargers to be operating and connected to the associated DC buses.
The LCO is modified by two Notes. Note 1 indicates that Vital Battery V may be substituted for any of the required vital batteries. However, the fifth battery cannot be declared OPERABLE until it is connected electrically in place of another battery and it has satisfied applicable Surveillance Requirements. Note 2 has been added to indicate that the C-S DG and its associated DC subsystem may be substituted for any of the required DGs. However, the C-S DG and its associated DC subsystem cannot be declared OPERABLE until it is connected electrically in place of another DG, and it has satisfied applicable Surveillance Requirements.
(continued)
Watts Bar-Unit 1                            B 3.8-57                                      Revision 113


The DC sources satisfy Criterion 3 of the NRC Policy Statement. LCO  Four 125V vital DC electrical power subsystems, each vital subsystem channel consisting of a battery bank, associated battery charger and the corresponding control equipment and interconnecting cabling supplying power to the associated DC bus within the channel; and four DG DC electrical power subsystems each consisting of a battery, a dual battery charger assembly, and the corresponding control equipment and interconnecting cabling are required to be OPERABLE to ensure the availability of the required power to shut down the reactor and maintain it in a safe condition after an anticipated operational occurrence (A00) or a postulated DBA. Loss of any DC electrical power subsystem does not prevent the minimum safety function from being performed (Ref. 4).
80.
An OPERABLE vital DC electrical power subsystem requires all required batteries and respective chargers to be operating and connected to the associated DC buses. The LCO is modified by two Notes. Note 1 indicates that Vital Battery V may be substituted for any of the required vital batteries. However, the fifth battery cannot be declared OPERABLE until it is connected electrically in place of another battery and it has satisfied applicable Surveillance Requirements. Note 2 has been added to indicate that the C-S DG and its associated DC subsystem may be substituted for any of the required DGs. However, the C-S DG and its associated DC subsystem cannot be declared OPERABLE until it is connected electrically in place of another DG, and it has satisfied applicable Surveillance Requirements. 
Given the following conditions:
    -  A perturbation occurred on the 161kV transmission grid.
    -  During the perturbation, the control power supply breaker, for the CSST C load tap changer for the Y winding TRIPS OPEN.
    -  The following indications are observed on 0-ECB-3:
Which ONE of the following describes the operability of the offsite power supply AND how the CSST C - Y winding voltage will be maintained?
In accordance with T/S LCO 3.8.1, AC Sources - Operating, the offsite power supply described above ____(1)____ operable.
In accordance with 1-PI-OPS-1-500KV, Main Control Room Voltage Monitoring, the associated SDBD will be maintained above the MINIMUM voltage requirement by
____(2)____.
A.   (1)  IS (2)  placing the DG on the SDBD B.   (1)  IS (2)  notifying the Northeast Area Dispatcher (NEAD) to ensure that the 161kV transmission alignments are adequate C.    (1)  IS NOT (2)  placing the DG on the SDBD D.     (1)  IS NOT (2)  notifying the Northeast Area Dispatcher (NEAD) to ensure that the 161kV transmission alignments are adequate


Four 125V vital DC electrical power subsystems, each vital subsystem channelconsisting of a battery bank, associated battery charger and the corresponding control equipment and interconnecting cabling supplying power to the associated DC bus within the channel;An OPERABLE vital DC electrical power subsystem requires all required batteries and respective chargers to be operating and connected to the associated DCbuses.Notethatachargermustbeoperating andconnected.  
CORRECT ANSWER:                                                                  B DISTRACTOR ANALYSIS:
: 80. Given the following conditions: - A perturbation occurred on the 161kV transmission grid. - During the perturbation, the control power supply breaker, for the CSST "C" load tap changer for the "Y" winding TRIPS OPEN. - The following indications are observed on 0-ECB-3:     Which ONE of the following describes the operability of the offsite power supply AND how the CSST C - Y winding voltage will be maintained?
A. Incorrect: While it is correct that the offsite power source is currently operable, it is not correct that the emergency diesel generator would be placed on the SDBD to maintain it operable. It is plausible to believe this as if such were done, the voltage of the SDBD would be certainly maintained within limits. Also, it very reasonably seems counterintuitive that one would adjust the entire 161kV grid voltage to compensate for the needs of one generating plant but that is precisely the case. Normally, the load tap changers account for the daily fluctuations in grid voltage. However, upon the loss of a nuclear facilitys capability to adjust for this, dispatch will coordinate with the remaining generating plants to maintain grid voltage.
B. Correct: As seen in 1-PI-OPS-1-500KV, Main Control Room Voltage Monitoring, WHEN CSST tap changer(s) have been placed in any of the following alternative alignments: Common Station Service Transformer C or Load Tap Changer Loss of Power or De-energizedTHEN NOTIFY NEAD of the alternative alignment. The basis for this notification is seen in the note preceding this step: Technical Specification operability is maintained in alternate alignment configuration for CSST Load Tap Changers by ensuring transmission alignments (TRO-TO-SOP-30.130, Watts Bar Nuclear Plant Grid Operating Guide) are adequate to ensure minimum voltage requirements are met. Furthermore, one may refer to 0-SI-82-2, 8 hour Diesel Generator AC power source operability verification to learn that the allowable voltage range for the 6.9kV SDBDs is 6800 to 7260VAC.
Therefore, the offsite power supply remains operable.
C. Incorrect: Again, it is incorrect and yet plausible placing the EDG on the SDBD would ensure that the minimum voltage requirement was met and that the operability of the offsite source was not maintained.
D. Incorrect: While it is correct that a notification would ensure that the minimum voltage require was meet, it is not correct that the operability of the offsite supply was not maintained. It is plausible to believe such because the original design output of the plant required that any time that a tap changer be placed in manual or de-energized, that the affiliated offsite power source be declared inoperable. Additionally, one may believe that 7.08kV is outside of the nominally allowed band (e.g. if they assumed that a +/-100VAC tolerance existed) for the 6.9 kV shutdown board. Notice that the allowable voltage range of 6800-7260 VAC is 100 VAC less than 6900VAC and 360 VAC greater than 6900VAC. Therefore, a +/- 100VAC band would be plausible.


In accordance with T/S LCO 3.8.1, AC Sources - Operating, the offsite power supply described above ____(1)____ operable.
Question Number:        80 Tier:    1  Group:      1 K/A:    077 Generator Voltage and Electric Grid Disturbances AA2. Ability to determine and interpret the following as they apply to Generator Voltage and Electric Grid Disturbances:
AA2.07 Operational status of engineered safety features Importance Rating:      3.6 4.0 10 CFR Part 55:      (CFR: 41.5 and 43.5 / 45.5, 45.7, and 45.8) 10CFR55.43.b:        10 CFR 55.43(b)(2)
K/A Match:    K/A is matched because the applicant is required to identify the operability of an offsite power source when its compensatory measure for electric grid disturbances is de-energized. The applicant must then identify the correct method by which the minimum voltage requirement would be met.
1-PI-OPS-1-500KV, Main Control Room Voltage Monitoring Technical


In accordance with 1-PI-OPS-1-500KV, "Main Control Room Voltage Monitoring,"  the associated SDBD will be maintained above the MINIMUM voltage requirement by ____(2)____. A. (1) IS (2) placing the DG on the SDBD B. (1) IS (2) notifying the Northeast Area Dispatcher (NEAD) to ensure that the 161kV transmission alignments are adequate C. (1) IS NOT (2) placing the DG on the SDBD D. (1) IS NOT  (2) notifying the Northeast Area Dispatcher (NEAD) to ensure that the 161kV transmission alignments are adequate CORRECT ANSWER:B DISTRACTOR ANALYSIS:  A. Incorrect: While it is correct that the offsite power source is currently operable, it is not correct that the emergency diesel generator would be placed on the SDBD to maintain it operable. It is plausible to believe this as if such were done, the voltage of the SDBD would be certainly maintained within limits. Also, it very reasonably seems counterintuitive that one would adjust the entire 161kV grid voltage to compensate for the needs of one generating plant but that is precisely the case. Normally, the load tap changers account for the daily fluctuations in grid voltage. However, upon the loss of a nuclear facility's capability to adjust for this, dispatch will coordinate with the remaining generating plants to maintain grid voltage. B. Correct:  As seen in 1-PI-OPS-1-500KV, "Main Control Room Voltage Monitoring," "WHEN CSST tap changer(s) have been placed in any of the following alternative alignments: Common Station Service Transformer C or Load Tap Changer Loss of Power or De-energized-THEN NOTIFY NEAD of the alternative alignment."  The basis for this notification is seen in the note preceding this step:  "Technical Specification operability is maintained in alternate alignment configuration for CSST Load Tap Changers by ensuring transmission alignments (TRO-TO-SOP-30.130, Watts Bar Nuclear Plant Grid Operating Guide) are adequate to ensure minimum voltage requirements are met."  Furthermore, one may refer to 0-SI-82-2, "8 hour Diesel Generator AC power source operability verification" to learn that the allowable voltage range for the 6.9kV SDBDs is 6800 to 7260VAC. Therefore, the offsite power supply remains operable. C. Incorrect: Again, it is incorrect and yet plausible placing the EDG on the SDBD would ensure that the minimum voltage requirement was met and that the operability of the offsite source was not maintained. D. Incorrect: While it is correct that a notification would ensure that the minimum voltage require was meet, it is not correct that the operability of the offsite supply was not maintained. It is plausible to believe such because the original design output of the plant required that any time that a tap changer be placed in manual or de-energized, that the affiliated offsite power source be declared inoperable. Additionally, one may believe that 7.08kV is outside of the nominally allowed band (e.g. if they assumed that a +/-100VAC tolerance existed) for the 6.9 kV shutdown board. Notice that the allowable voltage range of 6800-7260 VAC is 100 VAC less than 6900VAC and 360 VAC greater than 6900VAC. Therefore, a +/- 100VAC band would be plausible.
==Reference:==
 
0-SI-82-2, 8 hour Diesel Generator AC power source operability verification T/S Basis for LCO 3.8.1 Proposed references to     None be provided:
Question Number: 80  Tier:  1 Group:  1 K/A: 077 Generator Voltage and Electric Grid Disturbances AA2. Ability to determine and interpret the following as they apply to Generator Voltage and Electric Grid Disturbances: AA2.07 Operational status of engineered safety features Importance Rating: 3.6  4.0
Learning Objective:         3-OT-SYS245A 11.DESCRIBE the following aspects of Technical Specifications and Technical Requirements for this system:
 
: a. The conditions and required actions with completion time of one hour or less
10 CFR Part 55: (CFR: 41.5 and 43.5 / 45.5, 45.7, and 45.8)  10CFR55.43.b: 10 CFR 55.43(b)(2)  K/A Match: K/A is matched because the applicant is required to identify the operability of an offsite power source when its compensatory measure for electric grid disturbances is de-energized. The applicant must then identify the correct method by which the minimum voltage requirement would be met. Technical
: b. The Limiting Conditions for Operation, Applicability, and Bases.
Cognitive Level:
Higher              X Lower Question Source:
New                  X Modified Bank Bank Question History:           New question for the 2015-301 NRC SRO Exam Comments:                  See the marked up Clarification Guidance for SRO-only Questions.


==Reference:==
WBN                      Main Control Room                1-PI-OPS-1-500KV Unit 1                   Voltage Monitoring               Rev. 0007 Page 7 of 13 5.1       Voltage Control Monitoring (continued)
1-PI-OPS-1-500KV, Main Control Room Voltage Monitoring 0-SI-82-2, 8 hour Diesel Generator AC power source operability verification T/S Basis for LCO 3.8.1 Proposed references to be provided: None  Learning Objective: 3-OT-SYS245A 11.DESCRIBE the following aspects of Technical Specifications and Technical Requirements for this system:
NOTE VARS are to be maintained in accordance with section 5.2.
: a. The conditions and required actions with completion time of one hour or less  b. The Limiting Conditions for Operation, Applicability, and Bases. Cognitive Level:      Higher X  Lower    Question Source:      New X  Modified Bank  Bank      Question History: New question for the 2015-301 NRC SRO Exam  Comments: See the marked up Clarification Guidance for SRO-only Questions.
C. IF 500kV voltage is high, THEN ENSURE Main Generator VARS are incoming.
WBN Unit 1 Main Control Room Voltage Monitoring 1-PI-OPS-1-500KV Rev. 0007 Page 7 of 13 5.1 Voltage Control Monitoring (continued)     NOTE VARS are to be maintained in accordance with section 5.2. C. IF 500kV voltage is high, THEN   ENSURE Main Generator VARS are incoming. D. IF 500kV voltage is low, THEN ENSURE Main Generator VARS are outgoing. NOTE Tap Changers are normally operated in auto but can be operated in manual at SRO discretion. Operation in manual is considered an alternate alignment with respect to the operating requirements and limitations imposed by the WBN grid operating guide. Technical Specification operability is maintained in alternate alignment configuration for CSST Load Tap Changers by ensuring transmission alignments (TRO-TO-SOP-30.130, Watts Bar Nuclear Plant Grid Operating Guide) are adequate to ensure minimum voltage requirements are met. NEAD shall be notified when the alternate alignments are planned, entered, and exited. [3] WHEN CSST tap changer(s) have been placed in any of the following alternative alignments:
D. IF 500kV voltage is low, THEN ENSURE Main Generator VARS are outgoing.
NOTE Tap Changers are normally operated in auto but can be operated in manual at SRO discretion. Operation in manual is considered an alternate alignment with respect to the operating requirements and limitations imposed by the WBN grid operating guide.
Technical Specification operability is maintained in alternate alignment configuration for CSST Load Tap Changers by ensuring transmission alignments (TRO-TO-SOP-30.130, Watts Bar Nuclear Plant Grid Operating Guide) are adequate to ensure minimum voltage requirements are met. NEAD shall be notified when the alternate alignments are planned, entered, and exited.
[3]   WHEN CSST tap changer(s) have been placed in any of the following alternative alignments:
* 6.9kV Common Board A or B Loads on Alternate Feeders
* 6.9kV Common Board A or B Loads on Alternate Feeders
* 480V Turbine Building Common Board A or B on Alternate Feeder
* 480V Turbine Building Common Board A or B on Alternate Feeder
* Common Station Service Transformer C or D Controls on Alternate Feeder
* Common Station Service Transformer C or D Controls on Alternate Feeder
* Common Station Service Transformer C or D Load Tap Changer Loss of Power or De-energized
* Common Station Service Transformer C or D Load Tap Changer Loss of Power or De-energized
* Common Station Service Transformer C or D Load Tap Changer in OFF or in Manual During Modes 1 - 4 THEN NOTIFY NEAD of the alternative alignment. [4] NOTIFY NEAD within 30 minutes when Main Generator Voltage Regulator is NOT in automatic. End of Section WBN Unit 0 8 HOUR DIESEL GENERATOR AC POWER SOURCE OPERABILITY VERIFICATION 0-SI-82-2 Rev. 0013 Page 6 of 31   Date ________     Initials     4.0 PREREQUISITE ACTIONS 4.1 Preliminary Actions [1] RECORD Start Date and Time on Surveillance Task Sheet. ________ 4.2 Approvals and Notifications [1] OBTAIN SM/SRO approval to perform this instruction on Surveillance Task Sheet. ________ 5.0 ACCEPTANCE CRITERIA A. Each qualified offsite power circuit has the correct breaker alignment and indicated power available. B. Each DG tested is capable of starting from standby condition or modified start and achieving steady state voltage of greater than or equal to 6800 Volts and less than or equal to 7260 Volts and frequency greater than or equal to 58.8 Hz and less than or equal to 61.2 Hz.
* Common Station Service Transformer C or D Load Tap Changer in OFF or in Manual During Modes 1 - 4 THEN NOTIFY NEAD of the alternative alignment.
AC Sources - Operating B 3.8.1 BASES (continued)   (continued)    Watts Bar-Unit 1 B 3.8-3      APPLICABLE The initial conditions of DBA and transient analyses in the SAFETY ANALYSES FSAR, Section 6 (Ref. 4) and Section 15 (Ref. 5), assume ESF systems are OPERABLE. The AC electrical power sources are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to ESF systems so that the fuel, Reactor Coolant System (RCS), and containment design limits are not exceeded. These limits are discussed in more detail in the Bases for Section 3.2, Power Distribution Limits; Section 3.4, Reactor Coolant System (RCS); and Section 3.6, Containment Systems. The OPERABILITY of the AC electrical power sources is consistent with the initial assumptions of the Accident analyses and is based upon meeting the design basis of the plant. This results in maintaining at least two DG's associated with one load group or one offsite circuit OPERABLE during Accident conditions in the event of:     a. An assumed loss of all offsite power or all onsite AC power; and     b. A worst case single failure. The AC sources satisfy Criterion 3 of NRC Policy Statement. LCO   Two qualified circuits between the Watts Bar Hydro 161 kV switchyard and the onsite Class 1E Electrical Power System and separate and independent DGs for each train ensure availability of the required power to shut down the reactor and maintain it in a safe shutdown condition after an anticipated operational occurrence (AOO) or a postulated DBA. Qualified offsite circuits are those that are described in the FSAR and are part ofthe licensing basis for the plant. Each offsite circuit must be capable of maintaining acceptable frequency and voltage, and accepting required loads during an accident, while connected to the 6.9 kV shutdown boards. Offsite power from the Watts Bar Hydro 161 kV switchyard to the onsite Class 1E distribution system is from two independent immediate access circuits. Each of the two circuits are routed from the switchyard through a 161 kV transmission line and 161 to 6.9 kV transformer (common station service transformers) to the onsite Class 1E distribution system. The medium voltage power system starts at the low-side of the common station service transformers.
[4]   NOTIFY NEAD within 30 minutes when Main Generator Voltage Regulator is NOT in automatic.
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 3 of 16  II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]: A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)] Some examples of SRO exam items for this topic include:
End of Section
 
WBN             8 HOUR DIESEL GENERATOR AC                 0-SI-82-2 Unit 0          POWER SOURCE OPERABILITY                  Rev. 0013 VERIFICATION                  Page 6 of 31 Date ________                                                                   Initials 4.0   PREREQUISITE ACTIONS 4.1   Preliminary Actions
[1]   RECORD Start Date and Time on Surveillance Task Sheet.               ________
4.2   Approvals and Notifications
[1]   OBTAIN SM/SRO approval to perform this instruction on Surveillance Task Sheet.                                             ________
5.0   ACCEPTANCE CRITERIA A. Each qualified offsite power circuit has the correct breaker alignment and indicated power available.
B. Each DG tested is capable of starting from standby condition or modified start and achieving steady state voltage of greater than or equal to 6800 Volts and less than or equal to 7260 Volts and frequency greater than or equal to 58.8 Hz and less than or equal to 61.2 Hz.
 
AC Sources - Operating B 3.8.1 BASES (continued)
APPLICABLE       The initial conditions of DBA and transient analyses in the SAFETY ANALYSES   FSAR, Section 6 (Ref. 4) and Section 15 (Ref. 5), assume ESF systems are OPERABLE. The AC electrical power sources are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to ESF systems so that the fuel, Reactor Coolant System (RCS), and containment design limits are not exceeded. These limits are discussed in more detail in the Bases for Section 3.2, Power Distribution Limits; Section 3.4, Reactor Coolant System (RCS); and Section 3.6, Containment Systems.
The OPERABILITY of the AC electrical power sources is consistent with the initial assumptions of the Accident analyses and is based upon meeting the design basis of the plant. This results in maintaining at least two DG's associated with one load group or one offsite circuit OPERABLE during Accident conditions in the event of:
: a.       An assumed loss of all offsite power or all onsite AC power; and
: b.       A worst case single failure.
The AC sources satisfy Criterion 3 of NRC Policy Statement.
LCO               Two qualified circuits between the Watts Bar Hydro 161 kV switchyard and the onsite Class 1E Electrical Power System and separate and independent DGs for each train ensure availability of the required power to shut down the reactor and One must use this            maintain it in a safe shutdown condition after an anticipated operational statement in                occurrence (AOO) or a postulated DBA.
addition to the 1-Qualified offsite circuits are those that are described in the FSAR and are part PI-OPS-500kV to              ofthe licensing basis for the plant.
determine the operability of the          Each offsite circuit must be capable of maintaining acceptable frequency and SDBD.                        voltage, and accepting required loads during an accident, while connected to the 6.9 kV shutdown boards.
Offsite power from the Watts Bar Hydro 161 kV switchyard to the onsite Class 1E distribution system is from two independent immediate access circuits. Each of the two circuits are routed from the switchyard through a 161 kV transmission line and 161 to 6.9 kV transformer (common station service transformers) to the onsite Class 1E distribution system. The medium voltage power system starts at the low-side of the common station service transformers.
(continued)
Watts Bar-Unit 1                              B 3.8-3
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]:
A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)]
Some examples of SRO exam items for this topic include:
* Reporting requirements when the maximum licensed thermal power output is exceeded.
* Reporting requirements when the maximum licensed thermal power output is exceeded.
* Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems, fire doors, etc.
* Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems, fire doors, etc.
* The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
* The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
* National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
* National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
* Processes for TS and FSAR changes. Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic. B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)] Some examples of SRO exam items for this topic include:
* Processes for TS and FSAR changes.
Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic.
B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]
Some examples of SRO exam items for this topic include:
* Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
* Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
* Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
* Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
* Knowledge of TS bases that are required to analyze TS required actions and terminology.
* Knowledge of TS bases that are required to analyze TS required actions and terminology.
* Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).
* Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of  1 hour action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.  
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of  1 hour action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. ROs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Page 3 of 16


SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. RO's are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 4 of 16 ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)         Above this line RO knowledge Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 5 of 16 Figure 1:  Screening for SRO-only linked to 10 CFR 55.43(b)(2) (Tech Specs)       
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)
RO knowledge Above this line Page 4 of 16


Can question be answered solely by knowing  1 hour TS/TRM Action? RO question YesNoCan question be answered solely by knowing the LCO/TRM information listed "above-the-line?" YesRO question NoCan question be answered solely by knowing the TS Safety Limits? YesRO question NoDoes the question involve one or more of the following for TS, TRM, or ODCM?
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)
(Tech Specs)
Can question be answered solely by knowing  1       Yes hour TS/TRM Action?                                             RO question No Can question be answered solely by knowing the       Yes LCO/TRM information listed above-the-line?                   RO question No The voltage requirements for the SDBDs are contained in the S.R.s Can question be answered solely by knowing the       Yes TS Safety Limits?                                               RO question No Does the question involve one or more of the following for TS, TRM, or ODCM?
* Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
* Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)
* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)                               Yes      SRO-only
* Knowledge of TS bases that is required to analyze TS required actions and terminology SRO-only question YesNoQuestion might not be linked to 10 CFR 55.43(b)(2) for SRO-only
* Knowledge of TS bases that is required to analyze TS                   question required actions and terminology No Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only Page 5 of 16
: 81. Given the following conditions: - A LOCA has occurred on Unit 1. ECA-1.1, Loss of RHR Sump Recirculation is in progress. - CNTMT Pressure is RISING  - The CNTMT Critical Safety Function IS ORANGE. Which ONE of the following describes the appropriate procedure selection AND operation of the Containment Spray Pumps? The US will ____(1)____ and direct the crew to operate the Containment Spray Pumps as described in ____(2)____. NOTE: 1-FR-Z.1, High Containment Pressure 1-ECA-1.1, Loss of RHR Sump Recirculation A. (1) REMAIN in 1-ECA-1.1 (2) 1-ECA-1.1 B. (1) TRANSITION to 1-FR-Z.1 (2) 1-ECA-1.1 C. (1) REMAIN in 1-ECA-1.1 (2) 1-FR-Z.1. High Containment Pressure D. (1) TRANSITION to 1-FR-Z.1 (2) 1-FR-Z.1. High Containment Pressure CORRECT ANSWER:B DISTRACTOR ANALYSIS:   A. Incorrect: Plausible because there is another instruction (1-ES-1.3) associated with the containment sump that does take precedent over the Orange Path condition (thus a transition would not be made from 1-ES-1.3) and 1-FR-Z.1 does provide for the direct operation of the Containment Spray Pumps. B. Correct: Correct: The transition to 1-FR-Z.1 is required due to the ORANGE path, but the Containment Spray Pumps are required to be operated in accordance with1- ECA-1.1 as identified in both 1-ECA-1.1 and 1-FR-Z.1. C. Incorrect: Incorrect: Plausible because there is another instruction (1-ES-1.3) associated with the containment sump that does take precedent over the Orange Path condition (thus a transition would not be made from 1-ES-1.3) and the Containment Spray Pumps are required to be operated in accordance with 1-ECA-1.1 as identified in both 1-ECA-1.1 and 1-FR-Z.1. D. Incorrect: Plausible because the transition to 1-FR-Z.1 is required due to the ORANGE path, and 1-FR-Z.1 does provide for the direct operation of the Containment Spray Pumps.
 
Question Number: 81 Tier:   1 Group:   1 K/A: E11 Loss of Emergency Coolant Recirculation EA2. Ability to determine and interpret the following as they apply to the (Loss of Emergency Coolant Recirculation) EA2.2 Adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments.
81.
Importance Rating: 3.4 4.2 10 CFR Part 55: (CFR: 43.5 / 45.13) 10CFR55.43.b: 10 CFR 55.43(b)(5) K/A Match: K/A is matched because the question requires interpreting the conditions and adhering to the appropriate conditions within the emergency procedures which are required by the facility's license. SRO because the question requires knowledge of the content of the procedures versus knowledge of the overall mitigation strategy or purpose as well as the assessment of plant conditions, then selecting the procedure with which to proceed.
Given the following conditions:
    - A LOCA has occurred on Unit 1.
    -  1-ECA-1.1, Loss of RHR Sump Recirculation is in progress.
    - CNTMT Pressure is RISING .
    - The CNTMT Critical Safety Function IS ORANGE.
Which ONE of the following describes the appropriate procedure selection AND operation of the Containment Spray Pumps?
The US will ____(1)____ and direct the crew to operate the Containment Spray Pumps as described in ____(2)____.
NOTE: 1-FR-Z.1, High Containment Pressure 1-ECA-1.1, Loss of RHR Sump Recirculation A.   (1)   REMAIN in 1-ECA-1.1 (2)   1-ECA-1.1 B.   (1)   TRANSITION to 1-FR-Z.1 (2)   1-ECA-1.1 C.   (1)   REMAIN in 1-ECA-1.1 (2)   1-FR-Z.1. High Containment Pressure D.   (1)   TRANSITION to 1-FR-Z.1 (2)   1-FR-Z.1. High Containment Pressure
 
CORRECT ANSWER:                                                                 B DISTRACTOR ANALYSIS:
A. Incorrect: Plausible because there is another instruction (1-ES-1.3) associated with the containment sump that does take precedent over the Orange Path condition (thus a transition would not be made from 1-ES-1.3) and 1-FR-Z.1 does provide for the direct operation of the Containment Spray Pumps.
B. Correct: Correct: The transition to 1-FR-Z.1 is required due to the ORANGE path, but the Containment Spray Pumps are required to be operated in accordance with1- ECA-1.1 as identified in both 1-ECA-1.1 and 1-FR-Z.1.
C. Incorrect: Incorrect: Plausible because there is another instruction (1-ES-1.3) associated with the containment sump that does take precedent over the Orange Path condition (thus a transition would not be made from 1-ES-1.3) and the Containment Spray Pumps are required to be operated in accordance with 1-ECA-1.1 as identified in both 1-ECA-1.1 and 1-FR-Z.1.
D. Incorrect: Plausible because the transition to 1-FR-Z.1 is required due to the ORANGE path, and 1-FR-Z.1 does provide for the direct operation of the Containment Spray Pumps.
 
Question Number:         81 Tier:     1   Group:       1 K/A:   E11 Loss of Emergency Coolant Recirculation EA2. Ability to determine and interpret the following as they apply to the (Loss of Emergency Coolant Recirculation)
EA2.2 Adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments.
Importance Rating:       3.4 4.2 10 CFR Part 55:         (CFR: 43.5 / 45.13) 10CFR55.43.b:           10 CFR 55.43(b)(5)
K/A Match:     K/A is matched because the question requires interpreting the conditions and adhering to the appropriate conditions within the emergency procedures which are required by the facility's license. SRO because the question requires knowledge of the content of the procedures versus knowledge of the overall mitigation strategy or purpose as well as the assessment of plant conditions, then selecting the procedure with which to proceed.
Technical  
Technical  


==Reference:==
==Reference:==
1-ECA-1.1, Loss of RHR Sump Recirculation 1-FR-Z.1, High Containment Pressure Proposed references to be provided: None  Learning Objective: 3-OT-FRZ0001 2. Discuss the reasons that ECA-1.1, Loss of RHR Sump Recirculation, is given priority over 1-FR-Z.1, High Containment Pressure for directing Containment Spray operation. Cognitive Level:     Higher X Lower     Question Source:     New   Modified Bank   Bank X     Question History: Bank question W/E11 EA2.2 81 which was last used on the 06/2011 WBN NRC exam. Comments: The question is SRO only as detailed in the distractor analysis. The question meets the general SRO only criteria of "Assessment of facility conditions an selection of appropriate procedures during normal, abnormal, and emergency situations."
1-ECA-1.1, Loss of RHR Sump Recirculation 1-FR-Z.1, High Containment Pressure Proposed references to       None be provided:
WBN Unit 1 & 2 User's Guide for Abnormal and Emergency  Operating Instructions 0-TI-12.04 Rev. 0000 Page 30 of 57 2.5.2 Conflicts in Rules of Priority (continued)     B. Entry into 1-ECA-0.0, Loss of Shutdown Power, because of the complete loss of both trains of shutdown boards is expected to be a rare occurrence. 1. When 1-ECA-0.0 is implemented, special considerations come into effect. a. None of the electrically powered safeguards equipment used to restore Critical Safety Functions is operable. b. None of the FRs can be implemented.  
Learning Objective:         3-OT-FRZ0001
: c. A NOTE at the beginning of instruction 1-ECA-0.0 states that "Status Trees should be monitored for information only. The FRs should NOT be implemented". 2. Once in 1-ECA-0.0, the operator is NOT allowed to transition to any other instruction until some form of power is restored to the shutdown boards and a transition step is reached. 3. Permission to implement the FRs is NOT granted until some initial status checks and actions are performed by the operator. C. Certain instructions take precedence over FRs because of their treatment of specific initiating events. 1. Normally, this precedence is identified in a CAUTION or NOTE at the beginning of the specific instruction. 2. 1-ECA-1.1, Loss of RHR Sump Recirculation, directs the operator to perform actions which are intended to conserve RWST level. a. 1-ECA-1.1 directs the operator to shutdown containment spray pumps based upon containment pressure. b. This guidance is in conflict with the guidance of 1-FR-Z.1 which directs the operator to maintain all containment spray pumps in service. c. The guidance of 1-ECA-1.1 takes priority over the guidance of 1-FR-Z.1. d. 1-FR-Z.1 contains a CAUTION at the beginning of the instruction to remind the operator of this conflict and its correct resolution.
: 2. Discuss the reasons that ECA-1.1, Loss of RHR Sump Recirculation, is given priority over 1-FR-Z.1, High Containment Pressure for directing Containment Spray operation.
WBN Unit 1 & 2 User's Guide for Abnormal and Emergency  Operating Instructions 0-TI-12.04 Rev. 0000 Page 31 of 57 2.5.2 Conflicts in Rules of Priority (continued)     3. 1-ECA-2.1, Uncontrolled Depressurization of All Steam Generators, addresses depressurization, loss of level and resultant feed flow reduction to all steam generators. a. This condition results in a RED priority on the Heat Sink Status Tree. b. A CAUTION statement appears at the beginning of 1-ECA-2.1 and 1-FR-H.1 to identify that 1-FR-H.1 should NOT be implemented if the reduced feed flow condition is under the control of the operator. 4. 1-ECA-0.0 addresses a complete loss of shutdown power during which the actions of a FR in effect could NOT be completed successfully. a. If a complete loss of shutdown power is experienced, transition to 1-ECA-0.0 is required. b. 1-ECA-0.1 and 1-ECA-0.2 contain a note at the point where normal FR implementation can resume. Status Tree conditions should be reevaluated after that point in the instruction. 5. 1-ES-1.3, Transfer to Containment Sump, maintains suction supply to ECCS pumps and injection flowpath to the core. a. If RWST level reaches the low level setpoint and auto swapover is actuated or required, transition to 1-ES-1.3 is appropriate. b. 1-ES-1.3 transfer sequence steps are identified by a number on the control board (e.g. #1) to ensure minimum flowpath prior to continuing with the instruction in effect. 1-ES-1.3 should be implemented and completed through the transfer sequence (or transitioned from as directed in 1-ES-1.3). 2.5.3 Termination of EOI Usage A. EOI usage ends in one of the following ways with plant conditions stable: 1. Transition to a normal plant operating instruction, e.g., GOI. 2. On RHR System operation with COLD SHUTDOWN conditions. 3. On RHR System operation with either RHR containment sump recirculation or hot leg recirculation in service and with long term recovery actions being determined by the Technical Support Center.
Cognitive Level:
WBN Unit 1 High Containment Pressure 1-FR-Z.1 Rev. 0001    Page 3 of 7  Step Action/Expected Response Response Not Obtained 3.0 OPERATOR ACTIONS   CAUTION If 1-ECA-1.1, Loss of RHR Sump Recirculation, is in effect, the number of Cntmt spray pumps to be operated is directed in 1-ECA-1.1 rather than in this Instruction. NOTE Adverse containment setpoints [ADV] should be used where provided due to Phase B actuation. 1. ENSURE Cntmt spray operation: ESTABLISH at least one train of Cntmt spray flow. a. Cntmt spray signal ACTUATED.
Higher               X Lower Question Source:
: b. Cntmt spray pumps RUNNING.
New Modified Bank Bank                 X Question History:           Bank question W/E11 EA2.2 81 which was last used on the 06/2011 WBN NRC exam.
: c. Cntmt spray valves 1-FCV-72-2 and 1-FCV-72-39 OPEN. d. Cntmt spray pump suction valves OPEN:
Comments:                   The question is SRO only as detailed in the distractor analysis. The question meets the general SRO only criteria of Assessment of facility conditions an selection of appropriate procedures during normal, abnormal, and emergency situations.
* Valves from RWST: 1-FCV-72-21and 1-FCV-72-22 OR
 
* Valves from Cntmt sump: 1-FCV-72-44 and 1-FCV-72-45 e. Cntmt spray flow:
WBN                         User's Guide for                 0-TI-12.04 Unit 1 & 2              Abnormal and Emergency                Rev. 0000 Operating Instructions              Page 30 of 57 2.5.2   Conflicts in Rules of Priority (continued)
B. Entry into 1-ECA-0.0, Loss of Shutdown Power, because of the complete loss of both trains of shutdown boards is expected to be a rare occurrence.
: 1. When 1-ECA-0.0 is implemented, special considerations come into effect.
: a. None of the electrically powered safeguards equipment used to restore Critical Safety Functions is operable.
: b. None of the FRs can be implemented.
: c. A NOTE at the beginning of instruction 1-ECA-0.0 states that Status Trees should be monitored for information only. The FRs should NOT be implemented.
: 2. Once in 1-ECA-0.0, the operator is NOT allowed to transition to any other instruction until some form of power is restored to the shutdown boards and a transition step is reached.
: 3. Permission to implement the FRs is NOT granted until some initial status checks and actions are performed by the operator.
C. Certain instructions take precedence over FRs because of their treatment of specific initiating events.
: 1. Normally, this precedence is identified in a CAUTION or NOTE at the beginning of the specific instruction.
: 2. 1-ECA-1.1, Loss of RHR Sump Recirculation, directs the operator to perform actions which are intended to conserve RWST level.
: a. 1-ECA-1.1 directs the operator to shutdown containment spray pumps based upon containment pressure.
: b. This guidance is in conflict with the guidance of 1-FR-Z.1 which directs the operator to maintain all containment spray pumps in service.
: c. The guidance of 1-ECA-1.1 takes priority over the guidance of 1-FR-Z.1.
: d. 1-FR-Z.1 contains a CAUTION at the beginning of the instruction to remind the operator of this conflict and its correct resolution.
 
WBN                       User's Guide for                 0-TI-12.04 Unit 1 & 2              Abnormal and Emergency                Rev. 0000 Operating Instructions              Page 31 of 57 2.5.2     Conflicts in Rules of Priority (continued)
: 3. 1-ECA-2.1, Uncontrolled Depressurization of All Steam Generators, addresses depressurization, loss of level and resultant feed flow reduction to all steam generators.
: a. This condition results in a RED priority on the Heat Sink Status Tree.
: b. A CAUTION statement appears at the beginning of 1-ECA-2.1 and 1-FR-H.1 to identify that 1-FR-H.1 should NOT be implemented if the reduced feed flow condition is under the control of the operator.
: 4. 1-ECA-0.0 addresses a complete loss of shutdown power during which the actions of a FR in effect could NOT be completed successfully.
: a. If a complete loss of shutdown power is experienced, transition to 1-ECA-0.0 is required.
The basis for the plausibility for A &      b. 1-ECA-0.1 and 1-ECA-0.2 contain a note at the point where normal C.                              FR implementation can resume. Status Tree conditions should be reevaluated after that point in the instruction.
: 5. 1-ES-1.3, Transfer to Containment Sump, maintains suction supply to ECCS pumps and injection flowpath to the core.
: a. If RWST level reaches the low level setpoint and auto swapover is actuated or required, transition to 1-ES-1.3 is appropriate.
: b. 1-ES-1.3 transfer sequence steps are identified by a number on the control board (e.g. #1) to ensure minimum flowpath prior to continuing with the instruction in effect. 1-ES-1.3 should be implemented and completed through the transfer sequence (or transitioned from as directed in 1-ES-1.3).
2.5.3     Termination of EOI Usage A. EOI usage ends in one of the following ways with plant conditions stable:
: 1. Transition to a normal plant operating instruction, e.g., GOI.
: 2. On RHR System operation with COLD SHUTDOWN conditions.
: 3. On RHR System operation with either RHR containment sump recirculation or hot leg recirculation in service and with long term recovery actions being determined by the Technical Support Center.
 
WBN                 High Containment Pressure             1-FR-Z.1 Unit 1                                                    Rev. 0001 Step   Action/Expected Response                 Response Not Obtained 3.0     OPERATOR ACTIONS CAUTION       If 1-ECA-1.1, Loss of RHR Sump Recirculation, is in effect, the number of Cntmt spray pumps to be operated is directed in 1-ECA-1.1 rather than in this Instruction.
NOTE           Adverse containment setpoints [ADV] should be used where provided due to Phase B actuation.
: 1.     ENSURE Cntmt spray operation:             ESTABLISH at least one train of Cntmt spray flow.
: a. Cntmt spray signal ACTUATED.
: b. Cntmt spray pumps RUNNING.
: c. Cntmt spray valves 1-FCV-72-2 and 1-FCV-72-39 OPEN.
: d. Cntmt spray pump suction valves OPEN:
* Valves from RWST:
1-FCV-72-21and 1-FCV-72-22 OR
* Valves from Cntmt sump:
1-FCV-72-44 and 1-FCV-72-45
: e. Cntmt spray flow:
* 1-FI-72-34
* 1-FI-72-34
* 1-FI-72-13 WBN Unit 1 Loss of RHR Sump Recirculation 1-ECA-1.1 Rev. 0003  Step    Action/Expected Response   Response Not Obtained   Page 5 of 35  4. (continued) c. CHECK number of spray pumps running equal to number required. c. STOP and PULL TO LOCK any cntmt spray pump NOT required, AND CLOSE discharge valve(s) for pump(s) stopped:
* 1-FI-72-13 Page 3 of 7
 
WBN             Loss of RHR Sump Recirculation     1-ECA-1.1 Unit 1                                              Rev. 0003 Step   Action/Expected Response               Response Not Obtained
: 4.   (continued)
: c. CHECK number of spray pumps       c. STOP and PULL TO LOCK any running equal to number              cntmt spray pump NOT required, required.                            AND CLOSE discharge valve(s) for pump(s) stopped:
* 1-FCV-72-2 and/or
* 1-FCV-72-2 and/or
* 1-FCV-72-39 MANUALLY OPERATE spray pumps as required. DO NOT OPERATE cntmt spray pumps as required by FR-Z.1, High Containment Pressure, UNTIL either of the following:
* 1-FCV-72-39 MANUALLY OPERATE spray pumps as required.
DO NOT OPERATE cntmt spray pumps as required by FR-Z.1, High Containment Pressure, UNTIL either of the following:
* Cntmt spray pump suction aligned to cntmt sump, OR
* Cntmt spray pump suction aligned to cntmt sump, OR
* RWST makeup sufficient to support cntmt spray pump operation. WHEN cntmt sump level greater than 28%[36% ADV], THEN PERFORM steps 5, 6, and 7 as necessary. **GO TO Step 8 WBN Unit 1 & 2 User's Guide for Abnormal and Emergency  Operating Instructions 0-TI-12.04 Rev. 0000 Page 28 of 57 2.4.4 Status Tree Rules of Usage (continued)     C. Status Trees shall be monitored in the following priority: 1. 1-FR-S, Subcriticality,  
* RWST makeup sufficient to support cntmt spray pump operation.
: 2. 1-FR-C, Core Cooling,  
WHEN cntmt sump level greater than 28%[36% ADV], THEN PERFORM steps 5, 6, and 7 as necessary.
: 3. 1-FR-H, Heat Sink,  
                                                    **GO TO Step 8 Page 5 of 35
: 4. 1-FR-P, PTS,  
 
: 5. 1-FR-Z, Containment,  
WBN                         User's Guide for                 0-TI-12.04 Unit 1 & 2              Abnormal and Emergency                  Rev. 0000 Operating Instructions              Page 28 of 57 2.4.4   Status Tree Rules of Usage (continued)
: 6. 1-FR-I, Inventory. D. If a RED path is diagnosed, then the Function Restoration Instruction will be implemented IMMEDIATELY. E. If an ORANGE path is diagnosed, then the remaining Status Trees will be checked. If no RED path exits, then the highest priority ORANGE path Function Restoration Instruction will be implemented. F. Once implemented because of any RED or ORANGE path, that Function Restoration Instruction will be performed to completion or to a point of transition UNLESS a higher priority condition develops. 1. As a Function Restoration Instruction is performed, the status of that tree may change. This change does NOT change the priority of an instruction in progress. 2. If a higher priority condition develops, the instruction in effect should be suspended and the higher priority condition addressed. G. When no RED or ORANGE path exists, a YELLOW path Function Restoration Instruction can be implemented at the operator's discretion.
C. Status Trees shall be monitored in the following priority:
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 6 of 16  C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)] Some examples of SRO exam items for this topic include:
: 1. 1-FR-S, Subcriticality,
: 2. 1-FR-C, Core Cooling,
: 3. 1-FR-H, Heat Sink,
: 4. 1-FR-P, PTS,
: 5. 1-FR-Z, Containment,
: 6. 1-FR-I, Inventory.
D. If a RED path is diagnosed, then the Function Restoration Instruction will be implemented IMMEDIATELY.
E. If an ORANGE path is diagnosed, then the remaining Status Trees will be checked. If no RED path exits, then the highest priority ORANGE path Function Restoration Instruction will be implemented.
F. Once implemented because of any RED or ORANGE path, that Function Restoration Instruction will be performed to completion or to a point of transition UNLESS a higher priority condition develops.
: 1. As a Function Restoration Instruction is performed, the status of that tree may change. This change does NOT change the priority of an instruction in progress.
: 2. If a higher priority condition develops, the instruction in effect should be suspended and the higher priority condition addressed.
G. When no RED or ORANGE path exists, a YELLOW path Function Restoration Instruction can be implemented at the operators discretion.
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]
Some examples of SRO exam items for this topic include:
* 10 CFR 50.59 screening and evaluation processes.
* 10 CFR 50.59 screening and evaluation processes.
* Administrative processes for temporary modifications.
* Administrative processes for temporary modifications.
* Administrative processes for disabling annunciators.
* Administrative processes for disabling annunciators.
* Administrative processes for the installation of temporary instrumentation.
* Administrative processes for the installation of temporary instrumentation.
* Processes for changing the plant or plant procedures. Section IV provides an example of a satisfactory SRO-only question related to this topic. D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.       [10 CFR 55.43(b)(4)] Some examples of SRO exam items for this topic include:
* Processes for changing the plant or plant procedures.
Section IV provides an example of a satisfactory SRO-only question related to this topic.
D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
[10 CFR 55.43(b)(4)]
Some examples of SRO exam items for this topic include:
* Process for gaseous/liquid release approvals, i.e., release permits.
* Process for gaseous/liquid release approvals, i.e., release permits.
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures. Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits. SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc. E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedure's overall mitigative strategy or purpose.  
Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.
 
SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.
The applicant's knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedure's content is required to correctly answer the written test item, for example:
E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.
The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:
Page 6 of 16


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 7 of 16
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
* Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
* Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
SRO-only knowledge should not be claimed for questions that can be answered solely using "systems knowledge"; e.g.:
SRO-only knowledge should not be claimed for questions that can be answered solely using systems knowledge; e.g.:
* how the system works.
* how the system works.
* system flow path.
* system flow path.
* component locations, etc. SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:
* component locations, etc.
SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:
* the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
* the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
* any AOP entry condition.
* any AOP entry condition.
* plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
* plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
* immediate operator actions of a procedure. Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.        
* immediate operator actions of a procedure.
Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.
Page 7 of 16
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)
(Assessment and selection of procedures)
Can the question be answered solely by knowing systems knowledge, i.e., how the system works,      Yes RO question flowpath, logic, component location?
No Can the question be answered solely by knowing immediate operator actions?                            Yes    RO question No Can the question be answered solely by knowing entry conditions for AOPs or plant parameters          Yes RO question that require direct entry to major EOPs?
No Can the question be answered solely by knowing the purpose, overall sequence of events, or            Yes RO question overall mitigative strategy of a procedure?
No Does the question require one or more of the following?
* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps                                            Yes        SRO-only
* Knowledge of diagnostic steps and decision points in the              question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 8 of 16  Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5) (Assessment and selection of procedures)      
82.
Given the following conditions:
    -  Unit 1 is at 100% power.
    -  1-SI-99-10-A is in progress.
Subsequently:
    -  Unit 1 inadvertently trips due to the maintenance activity.
    -  Control Rod H-4 is stuck at 215 steps withdrawn.
Which ONE of the following describes the appropriate response?
In accordance with ES-0.1, an immediate boration ____(1)____ required.
Assuming that a condition exists where an Immediate Boration IS required, the SRO will assign the responsibility for the performance of 1-AOI-34 to the ____(2)____ in accordance with 0-TI-12.04.
NOTE: 1-SI-99-10-A, 62 Day Functional Test of SSPS Train A and Reactor Trip Breaker A ES-0.1, Reactor Trip Response 1-AOI-34, Immediate Boration 0-TI-12.04, Users Guide for Abnormal and Emergency Operating Instructions A.    (1)   IS (2)   OAC B.    (1)    IS NOT (2)    OAC C.    (1)    IS (2)    BOP/CRO D.    (1)    IS NOT (2)   BOP/CRO


Can the question be answered solely by knowing "systems knowledge", i.e., how the system works, flowpath, logic, component location? RO question YesNoCan the question be answered solely by knowing immediate operator actions? YesCan the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?YesNoDoes the question require one or more of the following?
CORRECT ANSWER:                                                                   D DISTRACTOR ANALYSIS:
* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
A   Incorrect: As seen in step 6 of ES-0.1, Reactor Trip Response, the crew is to ENSURE
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps
.                all control and shutdown rods fully inserted: RPIs at bottom scale. If a control rod remained withdrawn, the response not obtained would be used. Such response is: IF two or more rods are NOT fully inserted, THEN INITIATE borationREFER TO 1-AOI-34, Immediate Boration. Therefore, it is not correct that an immediate boration would be required on account of one rod which failed to insert. It is plausible to believe such based upon two facts. Firstly, step 6 of ES-0.1 requires that the RNO be utilized whenever any rod is not fully inserted.
* Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
Therefore, the very construct of the step would lead one to believe that action would be required on the account of one rod remaining withdrawn; this is due to the fact that step 6 contains two checks: check if all rods are inserted and then, if not, check if two or more rods are not inserted. One could remember the first check and believe that the ultimate action (the boration) depended upon that verification. Secondly, common sense would dictate that a compensatory measure would be required at any time that a reactor trip failed to insert its full negative reactivity (i.e. the failure of a rod to insert).
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?NoNoYesQuestion might not be linked to 10 CFR 55.43(b)(5) for SRO-only NoRO question RO question RO question 
TI-12.04, Users Guide for Abnormal and Emergency Operating Instructions directs that When running an AOI concurrently with an EOIthe Unit Supervisor/SRO will assign the BOP/CRO operator responsibility for the AOI.
: 82. Given the following conditions: - Unit 1 is at 100% power. SI-99-10-A is in progress. Subsequently: - Unit 1 inadvertently trips due to the maintenance activity. - Control Rod H-4 is stuck at 215 steps withdrawn. Which ONE of the following describes the appropriate response? In accordance with ES-0.1, an immediate boration ____(1)____ required. Assuming that a condition exists where an Immediate Boration IS required, the SRO will assign the responsibility for the performance of 1-AOI-34 to the ____(2)____ in accordance with 0-TI-12.04. A. (1) IS (2) OAC B. (1) IS NOT (2) OAC C. (1) IS  (2) BOP/CRO D. (1) IS NOT (2) BOP/CRO    NOTE: 1-SI-99-10-A, 62 Day Functional Test of SSPS Train A and Reactor Trip Breaker A  ES-0.1, Reactor Trip Response 1-AOI-34, Immediate Boration 0-TI-12.04, User's Guide for Abnormal and Emergency Operating Instructions CORRECT ANSWER:D DISTRACTOR ANALYSIS:   A. Incorrect: As seen in step 6 of ES-0.1, "Reactor Trip Response," the crew is to "ENSURE all control and shutdown rods fully inserted: RPIs at bottom scale."  If a control rod remained withdrawn, the response not obtained would be used. Such response is: "IF two or more rods are NOT fully inserted, THEN INITIATE boration-REFER TO 1-AOI-34, Immediate Boration."  Therefore, it is not correct that an immediate boration would be required on account of one rod which failed to insert. It is plausible to believe such based upon two facts. Firstly, step 6 of ES-0.1 requires that the RNO be utilized whenever any rod is not fully inserted. Therefore, the very construct of the step would lead one to believe that action would be required on the account of one rod remaining withdrawn; this is due to the fact that step 6 contains two checks: check if all rods are inserted and then, if not, check if two or more rods are not inserted. One could remember the first check and believe that the ultimate action (the boration) depended upon that verification. Secondly, common sense would dictate that a compensatory measure would be required at any time that a reactor trip failed to insert its full negative reactivity (i.e. the failure of a rod to insert).
Therefore, it is not correct that the OAC would be assigned the duty of 1-AOI-34.
TI-12.04, "User's Guide for Abnormal and Emergency Operating Instructions" directs that "When running an AOI concurrently with an EOI-the Unit Supervisor/SRO will assign the BOP/CRO operator responsibility for the AOI."  Therefore, it is not correct that the OAC would be assigned the duty of 1-AOI-34. It is very plausible to believe such as it is the normal duty of the OAC to initiate actions which directly affect the reactivity of the core (e.g. borate and/or dilute). Therefore, if the SRO had directly entered 1-AOI-34 and not passed such procedure off to an operator, he would direct the actions of such procedure to the OAC. Also, the next several steps of ES-0.1 monitor, control and initiate items which are normally under the responsibility of the BOP/CRO (e.g. the steam generators and secondary plant). Therefore, TI-12.04 requires that the OAC and BOP/CRO perform a "role swap" in this specific instance. B. Incorrect: As described, it is correct that the BOP/CRO would be assigned 1-AOI-34. However it is incorrect and yet plausible 1-AOI-34 would be implemented on account of one rod which failed to insert. C. Incorrect: While it is correct that 1-AOI-34 would be used whenever two or more rods remained withdrawn post reactor trip, it is not correct and yet plausible for reasons aforementioned that the OAC would perform 1-AOI-34. D. Correct: It is correct that 1-AOI-34 would be used whenever two or more rods remained withdrawn post reactor trip. Also, it is correct that in accordance with TI-12.04, the BOP/CRO would perform 1-AOI-34.  
It is very plausible to believe such as it is the normal duty of the OAC to initiate actions which directly affect the reactivity of the core (e.g. borate and/or dilute).
Therefore, if the SRO had directly entered 1-AOI-34 and not passed such procedure off to an operator, he would direct the actions of such procedure to the OAC. Also, the next several steps of ES-0.1 monitor, control and initiate items which are normally under the responsibility of the BOP/CRO (e.g. the steam generators and secondary plant). Therefore, TI-12.04 requires that the OAC and BOP/CRO perform a role swap in this specific instance.
B. Incorrect: As described, it is correct that the BOP/CRO would be assigned 1-AOI-34.
However it is incorrect and yet plausible 1-AOI-34 would be implemented on account of one rod which failed to insert.
C. Incorrect: While it is correct that 1-AOI-34 would be used whenever two or more rods remained withdrawn post reactor trip, it is not correct and yet plausible for reasons aforementioned that the OAC would perform 1-AOI-34.
D. Correct: It is correct that 1-AOI-34 would be used whenever two or more rods remained withdrawn post reactor trip. Also, it is correct that in accordance with TI-12.04, the BOP/CRO would perform 1-AOI-34.


Question Number: 82 Tier: 1 Group:   2 K/A: 005 Inoperable/Stuck Control Rod 2.4 Emergency Procedures / Plan 2.4.8 Knowledge of how abnormal operating procedures are used in conjunction with EOPs.
Question Number:       82 Tier:   1     Group:     2 K/A:   005 Inoperable/Stuck Control Rod 2.4 Emergency Procedures / Plan 2.4.8 Knowledge of how abnormal operating procedures are used in conjunction with EOPs.
Importance Rating: 3.8 4.5 10 CFR Part 55: (CFR: 41.10 / 43.5 / 45.13) 10CFR55.43.b: 10 CFR 55.43(b)(5) K/A Match: K/A is matched because the applicant is required to understand when and how 1-AOI-34, "Immediate Boration" is used in conjunction with ES-0.1, "Reactor Trip Response."  Such use occurs during the failure of control rod(s) to insert post reactor trip. Technical  
Importance Rating:     3.8 4.5 10 CFR Part 55:     (CFR: 41.10 / 43.5 / 45.13) 10CFR55.43.b:       10 CFR 55.43(b)(5)
K/A Match:   K/A is matched because the applicant is required to understand when and how 1-AOI-34, Immediate Boration is used in conjunction with ES-0.1, Reactor Trip Response. Such use occurs during the failure of control rod(s) to insert post reactor trip.
Technical             1-SI-99-10-A, 62 Day Functional Test of SSPS Train A and Reactor Trip Breaker A


==Reference:==
==Reference:==
1-SI-99-10-A, 62 Day Functional Test of SSPS Train A and Reactor Trip Breaker A ES-0.1, Reactor Trip Response 1-AOI-34, Immediate Boration TI-12.04, User's Guide for Abnormal and Emergency Operating Instructions   Proposed references to be provided: None  Learning Objective: 3-OT-AOI3400 8. DESCRIBE the reasons for the following responses as they apply to 1-AOI-34, Immediate Boration and the following: When emergency boration is required Actions contained in EOP for emergency boration   Cognitive Level:     Higher   Lower X   Question Source:     New X Modified Bank   Bank     Question History: New question for the 2015-301 NRC SRO Exam Comments: See the marked up Clarification Guidance for SRO-only Questions.
ES-0.1, Reactor Trip Response 1-AOI-34, Immediate Boration TI-12.04, Users Guide for Abnormal and Emergency Operating Instructions Proposed references to     None be provided:
WBN Unit 1 Reactor Trip Response ES-0.1 Rev. 0024    Page 7 of 21  Step Action/Expected Response Response Not Obtained 6. ENSURE all control and shutdown rods fully inserted:
Learning Objective:       3-OT-AOI3400
* RPIs at bottom scale. IF two or more rods are NOT fully inserted, THEN INITIATE boration of 3250 gals of greater than or equal to 6120 ppm boron for each rod not fully inserted:
: 8. DESCRIBE the reasons for the following responses as they apply to 1-AOI-34, Immediate Boration and the following:
* REFER TO AOI-34, Immediate Boration. 7. ANNOUNCE reactor trip over PA system. 8. MONITOR S/G levels: a. At least one S/G NR level greater than 29%. a. ENSURE feed flow greater than 410 gpm. b. S/G NR levels less than 50% and controlled. b. IF any S/G NR level continues to rise, THEN ISOLATE feed flow to affected S/G. 9. CONTROL S/G NR levels between 29% and 50%. 10. INITIATE BOP realignment:
When emergency boration is required Actions contained in EOP for emergency boration Cognitive Level:
Higher Lower               X Question Source:
New                 X Modified Bank Bank Question History:         New question for the 2015-301 NRC SRO Exam Comments:                 See the marked up Clarification Guidance for SRO-only Questions.
 
WBN                   Reactor Trip Response           ES-0.1 Unit 1                                                  Rev. 0024 Step   Action/Expected Response                 Response Not Obtained
: 6.     ENSURE all control and shutdown           IF two or more rods are NOT fully rods fully inserted:                     inserted, THEN
* RPIs at bottom scale.
INITIATE boration of 3250 gals of greater than or equal to 6120 ppm boron for each rod not fully inserted:
* REFER TO AOI-34, Immediate Boration.
: 7.     ANNOUNCE reactor trip over PA system.
: 8.     MONITOR S/G levels:
: a. At least one S/G NR level greater   a. ENSURE feed flow greater than than 29%.                                410 gpm.
: b. S/G NR levels less than 50% and     b. IF any S/G NR level continues to controlled.                              rise, THEN ISOLATE feed flow to affected S/G.
: 9.     CONTROL S/G NR levels between 29% and 50%.
: 10. INITIATE BOP realignment:
* REFER TO AOI-17, Turbine Trip.
* REFER TO AOI-17, Turbine Trip.
WBN Unit 1 & 2 User's Guide for Abnormal and Emergency  Operating Instructions 0-TI-12.04 Rev. 0000 Page 35 of 57 2.7 Prudent Operator Actions (continued)     3. The operator should consult nearby personnel who are suitably qualified and notify them of their proposed actions. If no disagreement is forthcoming, he should then take the necessary mitigation or preemptive actions to terminate the event. 4. The STAR principle should be applied --Stop, Think, Act, Review. Ask yourself: If I take this action, could I inadvertently cause other more severe problems? Am I better off taking no action at all? How will safety status be affected? 2.8 Use of AOIs While in EOIs 1. During performance of the 1-ES-0.1, if plant conditions warrant implementation of an AOI, then the required AOI may be performed concurrently (on a not-to-interfere basis) with the EOIs. 2. When running an AOI concurrently with an EOI (1-ECA-0.0, 1-ES-0.1, etc.) the Unit Supervisor/SRO will assign the BOP/CRO operator responsibility for the AOI if another Unit Supervisor is NOT available. If the BOP/CRO operator performs an AOI, he/she should consult directly with the Unit Supervisor and give them the status as required by the AOI. 3. When an AOI in effect directs a Reactor Trip, then the performance of the AOI should continue immediately following transition to 1-ES-0.1. Performance assignments will be at the discretion of the SM/US based on the status and importance of events in progress. 4. When implementing an AOI outside the "horseshoe" in the control room, the Unit Supervisor should accompany the board operator to read the procedure steps and direct actions of the operator, unless higher priority conditions demanding the Unit Supervisor's attention exist; in which case the BOP/CRO should implement the AOI using the single performer method. The actively licensed STA may serve as a reader unless the crew is in progress of performing actions within the EOI network. 3.0 RECORDS None Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 6 of 16  C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)] Some examples of SRO exam items for this topic include:
Page 7 of 21
 
WBN                       User's Guide for                 0-TI-12.04 Unit 1 & 2          Abnormal and Emergency                  Rev. 0000 Operating Instructions                Page 35 of 57 2.7     Prudent Operator Actions (continued)
: 3. The operator should consult nearby personnel who are suitably qualified and notify them of their proposed actions. If no disagreement is forthcoming, he should then take the necessary mitigation or preemptive actions to terminate the event.
: 4. The STAR principle should be applied --Stop, Think, Act, Review. Ask yourself: If I take this action, could I inadvertently cause other more severe problems? Am I better off taking no action at all? How will safety status be affected?
2.8     Use of AOIs While in EOIs
: 1. During performance of the 1-ES-0.1, if plant conditions warrant implementation of an AOI, then the required AOI may be performed concurrently (on a not-to-interfere basis) with the EOIs.
: 2. When running an AOI concurrently with an EOI (1-ECA-0.0, 1-ES-0.1, etc.)
the Unit Supervisor/SRO will assign the BOP/CRO operator responsibility for the AOI if another Unit Supervisor is NOT available. If the BOP/CRO operator performs an AOI, he/she should consult directly with the Unit Supervisor and give them the status as required by the AOI.
: 3. When an AOI in effect directs a Reactor Trip, then the performance of the AOI should continue immediately following transition to 1-ES-0.1.
Performance assignments will be at the discretion of the SM/US based on the status and importance of events in progress.
: 4. When implementing an AOI outside the horseshoe in the control room, the Unit Supervisor should accompany the board operator to read the procedure steps and direct actions of the operator, unless higher priority conditions demanding the Unit Supervisors attention exist; in which case the BOP/CRO should implement the AOI using the single performer method. The actively licensed STA may serve as a reader unless the crew is in progress of performing actions within the EOI network.
3.0     RECORDS None
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]
Some examples of SRO exam items for this topic include:
* 10 CFR 50.59 screening and evaluation processes.
* 10 CFR 50.59 screening and evaluation processes.
* Administrative processes for temporary modifications.
* Administrative processes for temporary modifications.
* Administrative processes for disabling annunciators.
* Administrative processes for disabling annunciators.
* Administrative processes for the installation of temporary instrumentation.
* Administrative processes for the installation of temporary instrumentation.
* Processes for changing the plant or plant procedures. Section IV provides an example of a satisfactory SRO-only question related to this topic. D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.       [10 CFR 55.43(b)(4)] Some examples of SRO exam items for this topic include:
* Processes for changing the plant or plant procedures.
Section IV provides an example of a satisfactory SRO-only question related to this topic.
D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
[10 CFR 55.43(b)(4)]
Some examples of SRO exam items for this topic include:
* Process for gaseous/liquid release approvals, i.e., release permits.
* Process for gaseous/liquid release approvals, i.e., release permits.
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures. Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits. SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc. E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedure's overall mitigative strategy or purpose.  
Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.
 
SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.
The applicant's knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedure's content is required to correctly answer the written test item, for example:
E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.
The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:
Page 6 of 16


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 7 of 16
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
* Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
* Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
SRO-only knowledge should not be claimed for questions that can be answered solely using "systems knowledge"; e.g.:
SRO-only knowledge should not be claimed for questions that can be answered solely using systems knowledge; e.g.:
* how the system works.
* how the system works.
* system flow path.
* system flow path.
* component locations, etc. SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:
* component locations, etc.
SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:
* the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
* the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
* any AOP entry condition.
* any AOP entry condition.
* plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
* plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
* immediate operator actions of a procedure. Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.        
* immediate operator actions of a procedure.
Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.
Page 7 of 16
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)
(Assessment and selection of procedures)
Can the question be answered solely by knowing systems knowledge, i.e., how the system works,      Yes RO question flowpath, logic, component location?
No Can the question be answered solely by knowing immediate operator actions?                            Yes    RO question No Can the question be answered solely by knowing entry conditions for AOPs or plant parameters          Yes RO question that require direct entry to major EOPs?
No Can the question be answered solely by knowing the purpose, overall sequence of events, or            Yes RO question overall mitigative strategy of a procedure?
No Does the question require one or more of the following?
* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps                                            Yes        SRO-only
* Knowledge of diagnostic steps and decision points in the              question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 8 of 16  Figure 2:  Screening for SRO-only linked to 10 CFR 55.43(b)(5) (Assessment and selection of procedures)       
83.
Given the following conditions:
    - 1-AOI-34, Immediate Boration, Section 3.2, Boration of RCS with CVCS in Service is in progress.
    -  1-HS-62-140A, VCT MAKEUP CONTROL has JUST been taken to START and released.
Subsequently:
    -  1-TANK-62-239, Boric Acid Tank A outlet piping RUPTURES.
    -  1-FI-62-139, BA TO BLENDER FLOW is 0 gpm.
    - 1-LI-62-238, BAT A LEVEL is RAPIDLY LOWERING .
Which ONE of the following completes the statement below?
In order to borate the RCS, the US must __________.
A. place the C BAT in service to the U1 CVCS blender in accordance with 1-SOI-62.05 B. continue in section 3.2 of 1-AOI-34 and align the RWST to the charging pump suction C. continue in section 3.2 of 1-AOI-34 and place the C BAT in service to the U1 CVCS blender D. alternate the charging pump suction to and from the RWST Using 1-LCV-62-135 and 136 in accordance with 1-SOI-62.01


Can the question be answered solely by knowing "systems knowledge", i.e., how the system works, flowpath, logic, component location? RO question YesNoCan the question be answered solely by knowing immediate operator actions? YesCan the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?YesNoDoes the question require one or more of the following?
CORRECT ANSWER:                                                                       B DISTRACTOR ANALYSIS:
* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
A. Incorrect: 1-AOI-34 will sequentially attempt methods of boration. Section 3.2 of such procedure will eventually direct the crew to place the Charging Pumps suction on the RWST. Therefore, 1-AOI-34 would provide a means of borating the RCS and because of this, the use of 1-SOI-62.05 is not required. This distractor is plausible because physically, it would be effective at providing a means of borating the RCS.
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps
B. Correct: Again, 1-AOI-34 would permit the alignment of the Charging Pump suctions to the RWST such that the RCS could be borated.
* Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
C Incorrect: 1-AOI-34 does not alternate which BAT is in service to the boric acid transfer pumps.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?NoNoYesQuestion might not be linked to 10 CFR 55.43(b)(5) for SRO-only NoRO question RO question RO question 
.            Additionally, it does not make any provision for placing an additional BAT in service. It is plausible to believe that it made such because that would be a very reasonable provision to be had in the event of a mechanical failure or low tank level. Additionally, if a very large boration were required (such as that required if multiple rods were stuck out after a reactor trip), the procedure does not directly provide for either BAT makeup or BAT swap (there are checks to validate that level is above that required by the T/Rs and subsequent REFER steps which direct the crew to an SOI).
: 83. Given the following conditions: AOI-34, Immediate Boration, Section 3.2, Boration of RCS with CVCS in Service is in progress. HS-62-140A, VCT MAKEUP CONTROL has JUST been taken to START and released. Subsequently: TANK-62-239, Boric Acid Tank 'A' outlet piping RUPTURES. FI-62-139, BA TO BLENDER FLOW is 0 gpm. LI-62-238, BAT A LEVEL is RAPIDLY LOWERING  Which ONE of the following completes the statement below? In order to borate the RCS, the US must __________. A. place the "C" BAT in service to the U1 CVCS blender in accordance with 1-SOI-62.05 B. continue in section 3.2 of 1-AOI-34 and align the RWST to the charging pump suction C. continue in section 3.2 of 1-AOI-34 and place the "C" BAT in service to the U1 CVCS blender D. alternate the charging pump suction to and from the RWST Using 1-LCV-62-135 and 136 in accordance with 1-SOI-62.01 CORRECT ANSWER:B DISTRACTOR ANALYSIS:   A. Incorrect: 1-AOI-34 will sequentially attempt methods of boration. Section 3.2 of such procedure will eventually direct the crew to place the Charging Pumps suction on the RWST. Therefore, 1-AOI-34 would provide a means of borating the RCS and because of this, the use of 1-SOI-62.05 is not required. This distractor is plausible because physically, it would be effective at providing a means of borating the RCS. B. Correct: Again, 1-AOI-34 would permit the alignment of the Charging Pump suctions to the RWST such that the RCS could be borated. C. Incorrect: 1-AOI-34 does not alternate which BAT is in service to the boric acid transfer pumps. Additionally, it does not make any provision for placing an additional BAT in service. It is plausible to believe that it made such because that would be a very reasonable provision to be had in the event of a mechanical failure or low tank level. Additionally, if a very large boration were required (such as that required if multiple rods were stuck out after a reactor trip), the procedure does not directly provide for either BAT makeup or BAT swap (there are checks to validate that level is above that required by the T/Rs and subsequent REFER steps which direct the crew to an SOI). D. Incorrect: The SOIs would not be used to provide an Immediate Boration given the conditions in the stem of this question. This particular distractor is plausible as it would provide a means of boration which is equivalent to that yielded by the AOI. Again, it is incorrect because the US is not required to utilize it (as 1-AOI-34 does provide a means of boration).
D Incorrect: The SOIs would not be used to provide an Immediate Boration given the conditions in the
Question Number: 83  Tier:  1 Group: 2 K/A: 024 Emergency Boration AA2. Ability to determine and interpret the following as they apply to the Emergency Boration:
.            stem of this question. This particular distractor is plausible as it would provide a means of boration which is equivalent to that yielded by the AOI. Again, it is incorrect because the US is not required to utilize it (as 1-AOI-34 does provide a means of boration).
AA2.04 Availability of BWST  Importance Rating: 3.4  4.2


10 CFR Part 55: (CFR: 43.5 / 45.13) 10CFR55.43.b: 10 CFR 55.43(b)(5) K/A Match: K/A is matched because the applicant must understand the impact of losing the Boric Acid Tank on an emergency boration and then correctly use the Emergency Boration procedure (1-AOI-34).
Question Number:        83 Tier:    1  Group:      2 K/A:    024 Emergency Boration AA2. Ability to determine and interpret the following as they apply to the Emergency Boration:
AA2.04 Availability of BWST Importance Rating:      3.4 4.2 10 CFR Part 55:       (CFR: 43.5 / 45.13) 10CFR55.43.b:         10 CFR 55.43(b)(5)
K/A Match:   K/A is matched because the applicant must understand the impact of losing the Boric Acid Tank on an emergency boration and then correctly use the Emergency Boration procedure (1-AOI-34).
Technical  
Technical  


==Reference:==
==Reference:==
1-AOI-34, Immediate Boration 0-SOI-62.05, Boric Acid Batching, Transfer, And Storage 1-SOI-62.01, CVCS-Charging and Letdown Proposed references to be provided: None  Learning Objective: 3-OT-AOI3400 9. DETERMINE the following as they apply to 1-AOI-34, Immediate Boration     Availability of Boric Acid Tanks Cognitive Level:     Higher X Lower     Question Source:     New X Modified Bank   Bank     Question History: New question for the 2015-301 NRC SRO Exam Comments: The question is SRO only as detailed in the distractor analysis. The question meets the general SRO only criteria of "Assessment of facility conditions an selection of appropriate procedures during normal, abnormal, and emergency situations."
1-AOI-34, Immediate Boration 0-SOI-62.05, Boric Acid Batching, Transfer, And Storage 1-SOI-62.01, CVCS-Charging and Letdown Proposed references to     None be provided:
WBN Unit 1 Immediate Boration 1-AOI-34 Rev. 0001 Page 4 of 24  3.0 OPERATOR ACTIONS 3.1 Diagnostics IF GO TO Subsection Page CVCS in service to RCS 3.2 5 CVCS shutdown or boration is required during Refueling 3.3 14 WBN Unit 1 Immediate Boration 1-AOI-34 Rev. 0001    Page 5 of 24  Step Action/Expected Response Response Not Obtained 3.2 Boration of RCS with CVCS in Service NOTE Boric acid addition should be noted to assist in determination of reactivity changes. 1. INITIATE normal boration to change CB as necessary: a. PLACE 1-HS-62-140B MODE SELECTOR, to BOR. b. CHECK 1-FC-62-139, BA TO BLENDER, indicates GPM. c. ADJUST 1-FC-62-139, BA TO BLENDER, setpoint to desired flow rate. d. ADJUST 1-FQ-62-139 BA BATCH COUNTER, to ensure boration continues. e. () MOMENTARILY PLACE 1-HS-62-140A, VCT MAKEUP CONTROL, to START and RELEASE. f. CHECK 1-HS-62-140A, Red light LIT. g. VERIFY boric acid flow indication on 1-FI-62-139, BA TO BLENDER FLOW. 2. ENSURE PW to blender isol, 1-FCV-62-143, CLOSED. 3. CHECK PW to blender flow, 1-FI-62-142, indicating ZERO. DISPATCH Operator to CLOSE PW to blender isolation, 1-ISV-62-933 [A4V/713].
Learning Objective:         3-OT-AOI3400
WBN Unit 1 Immediate Boration 1-AOI-34 Rev. 0001    3.2 Boration of RCS with CVCS in Service (continued)   Page 6 of 24  Step Action/Expected Response Response Not Obtained  NOTE A delay of 15 to 20 minutes may be expected before effects of negative reactivity insertion are observed. 4. MONITOR for negative reactivity insertion:
: 9. DETERMINE the following as they apply to 1-AOI-34, Immediate Boration Availability of Boric Acid Tanks Cognitive Level:
Higher               X Lower Question Source:
New                 X Modified Bank Bank Question History:           New question for the 2015-301 NRC SRO Exam Comments:                   The question is SRO only as detailed in the distractor analysis. The question meets the general SRO only criteria of Assessment of facility conditions an selection of appropriate procedures during normal, abnormal, and emergency situations.
 
WBN                   Immediate Boration               1-AOI-34 Unit 1                                                  Rev. 0001 3.0     OPERATOR ACTIONS The question 3.1     Diagnostics places the applicant in this section.
IF                                                   GO TO Subsection Page CVCS in service to RCS                                         3.2       5 CVCS shutdown or boration is required during                   3.3     14 Refueling Page 4 of 24
 
WBN                     Immediate Boration               1-AOI-34 Unit 1                                                    Rev. 0001 Step   Action/Expected Response                   Response Not Obtained 3.2     Boration of RCS with CVCS in Service NOTE           Boric acid addition should be noted to assist in determination of reactivity changes.
: 1.     INITIATE normal boration to change CB as necessary:
: a. PLACE 1-HS-62-140B MODE SELECTOR, to BOR.
: b. CHECK 1-FC-62-139, BA TO BLENDER, indicates GPM.
: c. ADJUST 1-FC-62-139, BA TO BLENDER, setpoint to desired                               The question stem flow rate.                                                 places the applicant at this
: d. ADJUST 1-FQ-62-139 BA                                     step.
BATCH COUNTER, to ensure boration continues.
: e.   () MOMENTARILY PLACE 1-HS-62-140A, VCT MAKEUP CONTROL, to START and RELEASE.                                             This is 0 due to the rupture.
: f. CHECK 1-HS-62-140A, Red light LIT.
: g. VERIFY boric acid flow indication on 1-FI-62-139, BA TO BLENDER FLOW.
: 2.     ENSURE PW to blender isol, 1-FCV-62-143, CLOSED.
: 3.     CHECK PW to blender flow,                   DISPATCH Operator to CLOSE PW to 1-FI-62-142, indicating ZERO.               blender isolation, 1-ISV-62-933
[A4V/713].
Page 5 of 24
 
WBN                   Immediate Boration                 1-AOI-34 Unit 1                                                    Rev. 0001 Step   Action/Expected Response                    Response Not Obtained 3.2     Boration of RCS with CVCS in Service (continued)
NOTE           A delay of 15 to 20 minutes may be expected before effects of negative reactivity insertion are observed.
Normal Boration will NOT be as it is going to the floor of
: 4.     MONITOR for negative reactivity             IF normal boration NOT insertingthe aux building.
insertion:                                 negative reactivity, THEN
* Neutron flux dropping.
* Neutron flux dropping.
                                                    ** GO TO Step 6.
* Tavg dropping.
* Tavg dropping.
* Control rod bank position rising (if in AUTO). IF normal boration NOT inserting negative reactivity, THEN 
* Control rod bank position rising (if in AUTO).
**  GO TO Step 6. 5. IF normal boration effective, THEN ** GO TO Step 9. 6. ESTABLISH required emergency boration flow: a. PLACE both BA pumps in FAST speed. b. () ADJUST emergency borate valve 1-FCV-62-138 to obtain required flow. b. () Locally ADJUST 1-FCV-62-138 to obtain required flow. c. CHECK emergency borate flow on 1-FI-62-137A. c. () Locally OPEN manual boration valve, 1-ISV-62-929 [Blender Station/713]. ENSURE BA flow control, 1-FCV-62-140, OPEN. ENSURE BA to Blender, 1-FI-62-139, indicating flow.
: 5.     IF normal boration effective, THEN This is impossible
WBN Unit 1 Immediate Boration 1-AOI-34 Rev. 0001    3.2 Boration of RCS with CVCS in Service (continued)   Page 7 of 24  Step Action/Expected Response Response Not Obtained  7. IF emergency boration flow established, THEN  
        ** GO TO Step 9.                                             as a Boric Acid Tank is still
** GO TO Step 9. 8. ALIGN RWST to CCP suction: a. () OPEN RWST outlet valves 1-LCV-62-135 and 1-LCV-62-136. [C.1] b. CLOSE VCT outlet valves 1-LCV-62-132 and 1-LCV-62-133. 9. REFER TO the following tech Specs:
: 6.     ESTABLISH required emergency                                 required to be boration flow:                                               aligned (the "A" BAT is still ruptured).
* 3.1.1, Shutdown Margin (SDM) - Tavg > 200°F.
: a. PLACE both BA pumps in FAST speed.
* 3.1.2, Shutdown Margin (SDM) - Tavg  200°F.
: b.   () ADJUST emergency borate           b.  () Locally ADJUST valve 1-FCV-62-138 to obtain               1-FCV-62-138 to obtain required required flow.                             flow.
: c. CHECK emergency borate flow           c.  () Locally OPEN manual boration on 1-FI-62-137A.                           valve, 1-ISV-62-929 [Blender Station/713].
ENSURE BA flow control, 1-FCV-62-140, OPEN.
ENSURE BA to Blender, 1-FI-62-139, indicating flow.
Page 6 of 24
 
WBN                   Immediate Boration             1-AOI-34 Unit 1                                                Rev. 0001 Step   Action/Expected Response                  Response Not Obtained 3.2     Boration of RCS with CVCS in Service (continued)
: 7.     IF emergency boration flow established, THEN
        ** GO TO Step 9.                                       This will provide the success path
: 8.     ALIGN RWST to CCP suction:                             for borating the RCS.
: a.   () OPEN RWST outlet valves 1-LCV-62-135 and 1-LCV-62-136. [C.1]
: b. CLOSE VCT outlet valves 1-LCV-62-132 and 1-LCV-62-133.
: 9.     REFER TO the following tech Specs:
* 3.1.1, Shutdown Margin (SDM) -
Tavg > 200°F.
* 3.1.2, Shutdown Margin (SDM) -
Tavg  200°F.
* 3.1.6, Shutdown Bank Insertion Limits.
* 3.1.6, Shutdown Bank Insertion Limits.
* 3.1.7, Control Bank Insertion Limits.
* 3.1.7, Control Bank Insertion Limits.
Line 447: Line 1,149:
* 3.5.4, Refueling Water Storage Tank (RWST).
* 3.5.4, Refueling Water Storage Tank (RWST).
* 3.9.1, Boron Concentration.
* 3.9.1, Boron Concentration.
WBN Unit 0 Boric Acid Batching, Transfer, And Storage 0-SOI-62.05 Rev. 0001 Page 41 of 124   Date________     Initials     8.2 Alternate BAT Operation 8.2.1 Place BAT C In Service With BA Pumps 1A & 1B Aligned to BAT C NOTE This Sect places BAT & BA Pumps in an alternate configuration, inconsistent with 1-TRI-62-3. [1] MOMENTARILY PLACE 1-HS-62-140A, VCT MAKEUP CONTROL [1-M-6], to STOP and RELEASE. ________ [2] CHECK 1-HS-62-140A Green light LIT. ________ [3] PERFORM the following: NOMENCLATURE LOCATION POSITION UNID PERF INITIAL BA PMP A 1-M-6 STOP 1-HS-62-230A   BA PMP B 1-M-6 STOP 1-HS-62-232A   [4] OPEN 1-ISV-62-1053B, BA XFER PUMP 1B-B DISCHARGE [A12R/713]. ________ [5] CLOSE the following valves: NOMENCLATURE LOCATION UNID PERF INITIAL BA XFER PUMP 1A-A RECIRC ISOL A12R/713 1-ISV-62-1054A BORIC ACID TANK A OUTLET A12R/713 1-ISV-62-1049 [6] OPEN 1-ISV-62-1048A, BA PUMP 1A-A/1B-B CROSSTIE [A12R/713]. ________
Page 7 of 24
WBN Unit 0 Boric Acid Batching, Transfer, And Storage 0-SOI-62.05 Rev. 0001 Page 42 of 124   Date________     Initials   8.2.1 Place BAT C In Service With BA Pumps 1A & 1B Aligned to BAT C (continued)     [7] IF Boric Acid Filter is bypassed, THEN ENSURE the following: A. CLOSE 1-ISV-62-1055A, BA XFER PUMP 1A-A BA FLTR A BYPASS. ________ B. OPEN 1-ISV-62-1055B, BA XFER PUMP 1B-B BA FLTR B BYPASS. ________ NOTE Both U1 BA Pumps are now aligned to BAT C. [8] START desired Boric Acid Pump (N/A pump NOT started): NOMENCLATURE LOCATION POSITION UNID PERF INITIAL BA PMP A 1-M-6 START 1-HS-62-230A   BA PMP B 1-M-6 START 1-HS-62-232A   [9] MOMENTARILY PLACE 1-HS-62-140A, VCT MAKEUP CONTROL [1-M-6], to START and RELEASE. ________ [10] CHECK 1-HS-62-140A, Red light LIT. ________ [11] WHEN desired to return to NORMAL alignment, THEN PERFORM the following: NOMENCLATURE LOCATION POSITION UNID PERF INITIAL VCT MAKEUP CONTROL 1-M-6 STOP 1-HS-62-140A BA PMP A 1-M-6 STOP 1-HS-62-230A   BA PMP B 1-M-6 STOP 1-HS-62-232A WBN Unit 0 Boric Acid Batching, Transfer, And Storage 0-SOI-62.05 Rev. 0001 Page 43 of 124   Date________     Initials   8.2.1 Place BAT C In Service With BA Pumps 1A & 1B Aligned to BAT C (continued)     [12] PERFORM the following: NOMENCLATURE LOCATION POSITION UNID PERF INITIAL VERIF INITIAL BA XFER PUMP 1B-B DISCHARGE A12R/713 CLOSED 1-ISV-62-1053B IVBA XFER PUMP 1A-A RECIRC ISOL A12R/713 OPEN 1-ISV-62-1054A IVBORIC ACID TANK A OUTLET A12R/713 OPEN 1-ISV-62-1049 IVBA PUMP 1A-A/1B-B CROSSTIE A12R/713 CLOSED 1-ISV-62-1048A IV[13] IF Boric Acid Filter is bypassed, THEN ENSURE the following: A. CLOSE 1-ISV-62-1055B, BA XFER PUMP 1B-B BA FLTR B BYPASS. ________ B. OPEN 1-ISV-62-1055A, BA XFER PUMP 1A-A BA FLTR A BYPASS. [14] START BA Pump 1A using 1-HS-62-230A, BA PMP A [1-M-6]. ________ [15] MOMENTARILY PLACE 1-HS-62-140A, VCT MAKEUP CONTROL [1-M-6], to START and RELEASE. ________ [16] CHECK 1-HS-62-140A, Red light LIT. ________ [17] IF BAT C is required to be in service, THEN PERFORM Sect 6.5, BAT C Normal Alignment. ________ End of Section WBN Unit 1 CVCS-Charging and Letdown 1-SOI-62.01 Rev. 0000 Page 82 of 108   Date ________     INITIALS     8.14 Alternating Charging Pump Suction To and From the RWST Using 1-LCV-62-135 And 136 [1] IF transferring Charging Pump Suction from the VCT to the RWST is desired, THEN PERFORM the following: [1.1] OPEN RWST to CVCS Charging Pump suction: NOMENCLATURE LOCATION UNID PERF INITIAL RWST TO CHARGING PMPS SUCTION 1-M-5 1-HS-62-135A RWST TO CHARGING PMPS SUCTION 1-M-5 1-HS-62-136A [1.2] CLOSE the following: NOMENCLATURE LOCATION UNID PERF INITIAL VCT TO CHARGING PMPS SUCTION 1-M-5 1-HS-62-132A VCT TO CHARGING PMPS SUCTION 1-M-5 1-HS-62-133A [1.3] ENSURE 1-FCV-62-1228 and 1-FCV-62-1229, CCP SUCTION TO VCT VENT HDR ISOL, CLOSED (green lights LIT). ________
 
WBN Unit 1 CVCS-Charging and Letdown 1-SOI-62.01 Rev. 0000 Page 83 of 108   Date ________     INITIALS   8.14 Alternating Charging Pump Suction To and From the RWST Using 1-LCV-62-135 And 136 (continued)     [2] IF transferring Charging Pump Suction from the RWST to the VCT is desired, THEN PERFORM the following: [2.1] ENSURE the following to align VCT to charging pumps: NOMENCLATURE LOCATION POSITION UNID PERF INITIAL VERIFIER INITIAL VCT TO CHARGING PMPS SUCTION 1-M-5 OPEN A-P AUTO1-HS-62-132A IVVCT TO CHARGING PMPS SUCTION 1-M-5 OPEN A-P AUTO1-HS-62-133A IV[2.2] ENSURE 1-FCV-62-1228 and 1-FCV-62-1229, CCP SUCTION TO VCT VENT HDR ISOL, OPEN (red lights LIT). ________ [2.3] ENSURE RWST to CVCS Charging Pump alignment: NOMENCLATURE LOCATION POSITION UNID PERF INITIAL VERIFIER INITIAL RWST TO CHARGING PMPS SUCTION 1-M-5 CLOSED A-P AUTO 1-HS-62-135A IVRWST TO CHARGING PMPS SUCTION 1-M-5 CLOSED A-P AUTO 1-HS-62-136A IV End of Section Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 6 of 16  C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)] Some examples of SRO exam items for this topic include:
WBN                       Boric Acid                 0-SOI-62.05 Unit 0          Batching, Transfer, And Storage      Rev. 0001 Page 41 of 124 Date________                                                                   Initials 8.2     Alternate BAT Operation                                             This alignment is physically possible.
8.2.1   Place BAT C In Service With BA Pumps 1A & 1B Aligned to             However, if the BATs were BAT C                                                               being credited for the Boration flow source/flow path, then a T/R impact would be had.
NOTE This Sect places BAT & BA Pumps in an alternate configuration, inconsistent with 1-TRI-62-3.
[1]   MOMENTARILY PLACE 1-HS-62-140A, VCT MAKEUP CONTROL [1-M-6], to STOP and RELEASE.                               ________
[2]   CHECK 1-HS-62-140A Green light LIT.                                 ________
[3]   PERFORM the following:
NOMENCLATURE       LOCATION     POSITION         UNID           PERF INITIAL BA PMP A               1-M-6       STOP       1-HS-62-230A BA PMP B               1-M-6       STOP       1-HS-62-232A
[4]   OPEN 1-ISV-62-1053B, BA XFER PUMP 1B-B DISCHARGE
[A12R/713].                                                         ________
[5]   CLOSE the following valves:
NOMENCLATURE           LOCATION               UNID               PERF INITIAL BA XFER PUMP 1A-A           A12R/713       1-ISV-62-1054A RECIRC ISOL BORIC ACID TANK A           A12R/713         1-ISV-62-1049 OUTLET
[6]   OPEN 1-ISV-62-1048A, BA PUMP 1A-A/1B-B CROSSTIE
[A12R/713].                                                         ________
 
WBN                           Boric Acid               0-SOI-62.05 Unit 0          Batching, Transfer, And Storage        Rev. 0001 Page 42 of 124 Date________                                                             Initials 8.2.1   Place BAT C In Service With BA Pumps 1A & 1B Aligned to BAT C (continued)
[7]   IF Boric Acid Filter is bypassed, THEN ENSURE the following:
A. CLOSE 1-ISV-62-1055A, BA XFER PUMP 1A-A BA FLTR A BYPASS.                                           ________
B. OPEN 1-ISV-62-1055B, BA XFER PUMP 1B-B BA FLTR B BYPASS.                                           ________
NOTE Both U1 BA Pumps are now aligned to BAT C.
[8]   START desired Boric Acid Pump (N/A pump NOT started):
NOMENCLATURE           LOCATION   POSITION       UNID       PERF INITIAL BA PMP A                   1-M-6     START     1-HS-62-230A BA PMP B                   1-M-6     START     1-HS-62-232A
[9]   MOMENTARILY PLACE 1-HS-62-140A, VCT MAKEUP CONTROL [1-M-6], to START and RELEASE.                         ________
[10] CHECK 1-HS-62-140A, Red light LIT.                             ________
[11] WHEN desired to return to NORMAL alignment, THEN PERFORM the following:
NOMENCLATURE           LOCATION   POSITION       UNID       PERF INITIAL VCT MAKEUP                 1-M-6     STOP       1-HS-62-140A CONTROL BA PMP A                   1-M-6     STOP       1-HS-62-230A BA PMP B                   1-M-6     STOP       1-HS-62-232A
 
WBN                           Boric Acid               0-SOI-62.05 Unit 0          Batching, Transfer, And Storage        Rev. 0001 Page 43 of 124 Date________                                                                 Initials 8.2.1   Place BAT C In Service With BA Pumps 1A & 1B Aligned to BAT C (continued)
[12]   PERFORM the following:
NOMENCLATURE           LOCATION     POSITION     UNID       PERF     VERIF INITIAL INITIAL BA XFER PUMP 1B-B           A12R/713     CLOSED 1-ISV-62-1053B DISCHARGE                                                                      IV BA XFER PUMP 1A-A           A12R/713     OPEN   1-ISV-62-1054A RECIRC ISOL                                                                    IV BORIC ACID TANK A           A12R/713     OPEN   1-ISV-62-1049 OUTLET                                                                          IV BA PUMP 1A-A/1B-B           A12R/713     CLOSED 1-ISV-62-1048A CROSSTIE                                                                        IV
[13]   IF Boric Acid Filter is bypassed, THEN ENSURE the following:
A. CLOSE 1-ISV-62-1055B, BA XFER PUMP 1B-B BA FLTR B BYPASS.                                                 ________
B. OPEN 1-ISV-62-1055A, BA XFER PUMP 1A-A BA FLTR A BYPASS.
[14]   START BA Pump 1A using 1-HS-62-230A, BA PMP A [1-M-6].             ________
[15]   MOMENTARILY PLACE 1-HS-62-140A, VCT MAKEUP CONTROL [1-M-6], to START and RELEASE.                             ________
[16]   CHECK 1-HS-62-140A, Red light LIT.                                 ________
[17]   IF BAT C is required to be in service, THEN PERFORM Sect 6.5, BAT C Normal Alignment.                           ________
End of Section
 
WBN                 CVCS-Charging and Letdown           1-SOI-62.01 Unit 1                                                  Rev. 0000 Page 82 of 108 Date ________                                                             INITIALS 8.14   Alternating Charging Pump Suction To and From the RWST Using 1-LCV-62-135 And 136
[1]     IF transferring Charging Pump Suction from the VCT to the RWST is desired, THEN PERFORM the following:
[1.1]     OPEN RWST to CVCS Charging Pump suction:
PERF NOMENCLATURE               LOCATION           UNID         INITIAL RWST TO CHARGING               1-M-5       1-HS-62-135A PMPS SUCTION RWST TO CHARGING               1-M-5       1-HS-62-136A PMPS SUCTION
[1.2]     CLOSE the following:
PERF NOMENCLATURE               LOCATION           UNID         INITIAL VCT TO CHARGING               1-M-5       1-HS-62-132A PMPS SUCTION VCT TO CHARGING               1-M-5       1-HS-62-133A PMPS SUCTION
[1.3]     ENSURE 1-FCV-62-1228 and 1-FCV-62-1229, CCP SUCTION TO VCT VENT HDR ISOL, CLOSED (green lights LIT).                                             ________
 
WBN                 CVCS-Charging and Letdown             1-SOI-62.01 Unit 1                                                    Rev. 0000 Page 83 of 108 Date ________                                                                 INITIALS 8.14   Alternating Charging Pump Suction To and From the RWST Using 1-LCV-62-135 And 136 (continued)
[2]     IF transferring Charging Pump Suction from the RWST to the VCT is desired, THEN PERFORM the following:
[2.1]     ENSURE the following to align VCT to charging pumps:
PERF    VERIFIER NOMENCLATURE           LOCATION   POSITION         UNID       INITIAL   INITIAL VCT TO CHARGING           1-M-5     OPEN       1-HS-62-132A PMPS SUCTION                      A-P AUTO                                        IV VCT TO CHARGING           1-M-5     OPEN       1-HS-62-133A PMPS SUCTION                      A-P AUTO                                        IV
[2.2]     ENSURE 1-FCV-62-1228 and 1-FCV-62-1229, CCP SUCTION TO VCT VENT HDR ISOL, OPEN (red lights LIT).                                       ________
[2.3]     ENSURE RWST to CVCS Charging Pump alignment:
PERF    VERIFIER NOMENCLATURE           LOCATION   POSITION         UNID       INITIAL   INITIAL RWST TO CHARGING           1-M-5     CLOSED       1-HS-62-135A PMPS SUCTION                      A-P AUTO                                        IV RWST TO CHARGING           1-M-5     CLOSED       1-HS-62-136A PMPS SUCTION                      A-P AUTO                                        IV End of Section
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]
Some examples of SRO exam items for this topic include:
* 10 CFR 50.59 screening and evaluation processes.
* 10 CFR 50.59 screening and evaluation processes.
* Administrative processes for temporary modifications.
* Administrative processes for temporary modifications.
* Administrative processes for disabling annunciators.
* Administrative processes for disabling annunciators.
* Administrative processes for the installation of temporary instrumentation.
* Administrative processes for the installation of temporary instrumentation.
* Processes for changing the plant or plant procedures. Section IV provides an example of a satisfactory SRO-only question related to this topic. D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.       [10 CFR 55.43(b)(4)] Some examples of SRO exam items for this topic include:
* Processes for changing the plant or plant procedures.
Section IV provides an example of a satisfactory SRO-only question related to this topic.
D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
[10 CFR 55.43(b)(4)]
Some examples of SRO exam items for this topic include:
* Process for gaseous/liquid release approvals, i.e., release permits.
* Process for gaseous/liquid release approvals, i.e., release permits.
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures. Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits. SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc. E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedure's overall mitigative strategy or purpose.  
Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.
 
SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.
The applicant's knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedure's content is required to correctly answer the written test item, for example:
E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.
The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:
Page 6 of 16


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 7 of 16
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
* Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
* Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
SRO-only knowledge should not be claimed for questions that can be answered solely using "systems knowledge"; e.g.:
SRO-only knowledge should not be claimed for questions that can be answered solely using systems knowledge; e.g.:
* how the system works.
* how the system works.
* system flow path.
* system flow path.
* component locations, etc. SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:
* component locations, etc.
SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:
* the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
* the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
* any AOP entry condition.
* any AOP entry condition.
* plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
* plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
* immediate operator actions of a procedure. Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.        
* immediate operator actions of a procedure.
 
Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 8 of 16   Figure 2:  Screening for SRO-only linked to 10 CFR 55.43(b)(5) (Assessment and selection of procedures)       
Page 7 of 16


Can the question be answered solely by knowing "systems knowledge", i.e., how the system works, flowpath, logic, component location? RO question YesNoCan the question be answered solely by knowing immediate operator actions? YesCan the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?YesNoDoes the question require one or more of the following?
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps
(Assessment and selection of procedures)
* Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
Can the question be answered solely by knowing systems knowledge, i.e., how the system works,       Yes RO question flowpath, logic, component location?
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?NoNoYesQuestion might not be linked to 10 CFR 55.43(b)(5) for SRO-only NoRO question RO question RO question 
No                                                    The question cannot be answered solely on systems Can the question be answered solely by knowing                     knowledge because such immediate operator actions?                           Yes    RO question knowledge would cause the No                                                    applicant to simultaneously arrive at two answers (because knowledge of the Can the question be answered solely by knowing                     procedures is required).
: 84. Given the following timeline: 00:00:00 Unit 1 is in 
entry conditions for AOPs or plant parameters         Yes RO question that require direct entry to major EOPs?
No Can the question be answered solely by knowing the purpose, overall sequence of events, or            Yes RO question overall mitigative strategy of a procedure?
No The question is asking details of Does the question require one or more of the following?             the procedure beyond the overall
* Assessing plant conditions (normal, abnormal, or               strategy.
emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps                                           Yes        SRO-only
* Knowledge of diagnostic steps and decision points in the               question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant The question is normal, abnormal, and emergency procedures asking whether one No                                                        must select an SOI to provide a success path or Question might not be linked to                           proceed with the 10 CFR 55.43(b)(5) for SRO-only AOI in use.
Page 8 of 16


The shutdown rods are FULLY WITHDRAWN. 00:0 1 :00 Source Range Nuclear Instrument N-31 fails LOW. The crew implements 1-AOI-4, Nuclear Instrumentation Malfunctions. 00: 10 : 00 Source Range Nuclear Instrument N-32 fails LOW. The OAC takes 1-RT-1, REACTOR TRIP to TRIP.
84.
The US enters T/S LCO 3.3.1 condition L: Which ONE of the following completes the statements below? In accordance with 0-TI-12.04, 1-E-0 ____(1)____ be entered to confirm the reactor trip. In accordance with the Unit 1 T/S, SR 3.1.1.1 MUST be completed by _____(2)_____. A. (1) MUST (2) 0 1: 10 : 00 B. (1) MUST (2) 0 1:25:00 C. (1) NEED NOT (2) 0 1: 10 : 00 D. (1) NEED NOT (2) 0 1:25:00 NOTE: 1-E-0, Reactor Trip or Safety Injection  T/S LCO, 3.3.1 Reactor Trip System (RTS) Instrumentation  0-TI-12.04, User's Guide for Abnormal and Emergency Operating Instructions CORRECT ANSWER:C DISTRACTOR ANALYSIS:  A. Incorrect: 0-TI-12.04, demonstrates that the EOI network contains implementation points which are applicable in Modes 1,2,3 or 4. Therefore, it is incorrect that with the Unit in Mode 5, that the EOI network would be implemented. It is plausible to believe this because there is no procedure other than 1-E-0 which is written to address a "reactor trip response" and that the stem of the question presents the applicant with a reactor trip.
Given the following timeline:
00:00:00 Unit 1 is in The shutdown rods are FULLY WITHDRAWN.
00:0 1 :00 Source Range Nuclear Instrument N-31 fails LOW.
The crew implements 1-AOI-4, Nuclear Instrumentation Malfunctions.
00: 10 : 00 Source Range Nuclear Instrument N-32 fails LOW.
The OAC takes 1-RT-1, REACTOR TRIP to TRIP.
The US enters T/S LCO 3.3.1 condition L:
Which ONE of the following completes the statements below?
In accordance with 0-TI-12.04, 1-E-0 ____(1)____ be entered to confirm the reactor trip.
In accordance with the Unit 1 T/S, SR 3.1.1.1 MUST be completed by _____(2)_____.
NOTE: 1-E-0, Reactor Trip or Safety Injection T/S LCO, 3.3.1 Reactor Trip System (RTS) Instrumentation 0-TI-12.04, Users Guide for Abnormal and Emergency Operating Instructions A.     (1)   MUST (2)   0 1: 10 : 00 B.     (1)   MUST (2)   0 1:25:00 C.     (1)   NEED NOT (2)   0 1: 10 : 00 D.     (1)   NEED NOT (2)   0 1:25:00


SR 3.0.2 states that "If a Completion Time requires periodic performance on a "once per . . ." basis, the above Frequency extension applies to each performance after the initial performance."  Therefore, the first performance of SR 3.1.1.1 is required 1 hour after the T/S LCO required action entry time or at 0110. B. Incorrect: Again it is incorrect and yet plausible that the "Reactor Trip Response" procedure would be used following a reactor trip in Mode 5. It is also incorrect that the first performance of SR 3.1.1.1 would be required at 0125. It is plausible to believe this as SR 3.0.2 does provide that: "The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency."  However, as mentioned, this is only applicable to performances conducted after the initial. C. Correct: It is correct that 1-E-0 need not be entered in Mode 5 to confirm a reactor trip. It is also correct that the first performance of the surveillance is required at 0110. D. Incorrect: While it is correct that 1-E-0 need not be entered in Mode 5 to confirm a reactor trip, it is incorrect and yet plausible that the first performance of the surveillance is required at 0125.
CORRECT ANSWER:                                                                  C DISTRACTOR ANALYSIS:
Question Number: 84  Tier:  1 Group:  2 K/A: 032 Loss of Source Range Nuclear Instrumentation AA2. Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation: AA2.06 Confirmation of reactor trip Importance Rating: 3.9  4.1
A. Incorrect: 0-TI-12.04, demonstrates that the EOI network contains implementation points which are applicable in Modes 1,2,3 or 4.
Therefore, it is incorrect that with the Unit in Mode 5, that the EOI network would be implemented. It is plausible to believe this because there is no procedure other than 1-E-0 which is written to address a reactor trip response and that the stem of the question presents the applicant with a reactor trip.
SR 3.0.2 states that If a Completion Time requires periodic performance on a "once per . . ." basis, the above Frequency extension applies to each performance after the initial performance.
Therefore, the first performance of SR 3.1.1.1 is required 1 hour after the T/S LCO required action entry time or at 0110.
B. Incorrect: Again it is incorrect and yet plausible that the Reactor Trip Response procedure would be used following a reactor trip in Mode 5. It is also incorrect that the first performance of SR 3.1.1.1 would be required at 0125. It is plausible to believe this as SR 3.0.2 does provide that:
The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency.
However, as mentioned, this is only applicable to performances conducted after the initial.
C. Correct: It is correct that 1-E-0 need not be entered in Mode 5 to confirm a reactor trip. It is also correct that the first performance of the surveillance is required at 0110.
D. Incorrect: While it is correct that 1-E-0 need not be entered in Mode 5 to confirm a reactor trip, it is incorrect and yet plausible that the first performance of the surveillance is required at 0125.


10 CFR Part 55: (CFR: 43.5 / 45.13) 10CFR55.43.b: 10 CFR 55.43(b)(2) and 10 CFR 55.43(b)(5) K/A Match: K/A is matched because while injected into a loss of source range instrumentation, the applicant is required to determine if 1-E-0 is required to confirm a reactor trip in Mode 5. The second part of the question requires the applicant to determine when the first performance of a shutdown margin verification would be required post reactor trip.
Question Number:        84 Tier:    1  Group:        2 K/A:    032 Loss of Source Range Nuclear Instrumentation AA2. Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation:
AA2.06 Confirmation of reactor trip Importance Rating:      3.9 4.1 10 CFR Part 55:       (CFR: 43.5 / 45.13) 10CFR55.43.b:         10 CFR 55.43(b)(2) and 10 CFR 55.43(b)(5)
K/A Match:   K/A is matched because while injected into a loss of source range instrumentation, the applicant is required to determine if 1-E-0 is required to confirm a reactor trip in Mode 5. The second part of the question requires the applicant to determine when the first performance of a shutdown margin verification would be required post reactor trip.
Technical  
Technical  


==Reference:==
==Reference:==
T/S LCO 3.3.1, Reactor Trip System (RTS) Instrumentation 0-TI-12.04, User's Guide for Abnormal and Emergency Operating Instructions Proposed references to be provided: None  Learning Objective: 3-OT-SYS092A 15. Given a set of plant conditions/parameters, APPLY the appropriate Technical Specifications and Technical Requirements.
T/S LCO 3.3.1, Reactor Trip System (RTS) Instrumentation 0-TI-12.04, Users Guide for Abnormal and Emergency Operating Instructions Proposed references to     None be provided:
3-OT-EOP0000  
Learning Objective:         3-OT-SYS092A
: 8. Analyze a set of plant conditions and identify required procedure transitions  
: 15. Given a set of plant conditions/parameters, APPLY the appropriate Technical Specifications and Technical Requirements.
: 15. Explain the purpose for and the basis of each step in E-0   Cognitive Level:     Higher   Lower X   Question Source:     New X Modified Bank   Bank     Question History: New question for the 2015-301 NRC SRO Exam Comments:
3-OT-EOP0000
WBN Unit 1 & 2 User's Guide for Abnormal and Emergency  Operating Instructions 0-TI-12.04 Rev. 0000 Page 7 of 57   2.1.2 Mode Applicability of the EOIs The EOIs are written to mitigate emergency transients initiated when the unit is at "hot" or "power" conditions. A. The guidance for operator action in the EOIs assumes that the safety-related equipment required by Tech Specs in Mode 1 or Mode 2 is available for use. B. The operating crew should implement the EOI network whenever reactor trip or safety injection events are initiated with the unit in Modes 1, 2, or 3. C. The operating crew should implement the EOI network for the complete loss of shutdown power event with the unit in Modes 1, 2, 3, or 4. D. Implementation of the EOI network in Mode 4 requires the operating crew to consider plant conditions and each specific instruction's applicability. 1. The EOI network assumes that the Residual Heat Removal (RHR) system is aligned for its Emergency Core Cooling mode. 2. Although most of the FRs can be utilized to respond to events during Mode 4 conditions, they assume ECCS equipment has operated and steam generators are available and required for heat removal. 3. Events (other than complete loss of shutdown power) initiated with the unit in Modes 4, 5, or 6 should be mitigated by implementation of the Abnormal Operating Instructions (AOIs). 4. The operating team should consider implementation of the EOI network if events initiated in Modes 4 or 5 result in plant heat-up to Mode 3. 5. A specific task (e.g., alignment of RHR sump recirc to SI pump suction) that is detailed in the EOIs may be appropriate during an unanticipated event. When considering such actions, the crew must be cautious and NOT apply the instruction out of context.
: 8. Analyze a set of plant conditions and identify required procedure transitions
Frequency 1.4 1.4 Frequency   (continued)      Watts Bar-Unit 1 1.4-3      EXAMPLES EXAMPLE 1.4-2   (continued)   SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify flow is within limits. Once within 12 hours after 25% RTP AND 24 hours thereafter       Example 1.4-2 has two Frequencies. The first is a one time performance Frequency, and the second is of the type shown in Example 1.4-1. The logical connector "AND" indicates that both Frequency requirements must be met. Each time reactor power is increased from a power level < 25% RTP to 25% RTP, the Surveillance must be performed within 12 hours. The use of "once" indicates a single performance will satisfy the specified Frequency (assuming no other Frequencies are connected by "AND"). This type of Frequency does not qualify for the extension allowed by SR 3.0.2. "Thereafter" indicates future performances must be established per SR 3.0.2, but only after a specified condition is first met (i.e., the "once" performance in this example). If reactor power decreases to < 25% RTP, the measurement of both intervals stops. New intervals start upon reactor power reaching 25% RTP.
: 15. Explain the purpose for and the basis of each step in E-0 Cognitive Level:
SR Applicability 3.0      (continued)   Watts Bar-Unit 1 3.0-4 Amendment 42      3.0  SURVEILLANCE REQUIREMENT (SR) APPLICABILITY  SR  3.0.1  SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3. Surveillances do not have to be performed on inoperable equipment or variables outside specified limits. SR  3.0.2  The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met. For Frequencies specified as "once," the above interval extension does not apply.
Higher Lower               X Question Source:
If a Completion Time requires periodic performance on a "once per . . ." basis, the above Frequency extension applies to each performance after the initial performance.
New                 X Modified Bank Bank Question History:           New question for the 2015-301 NRC SRO Exam Comments:
 
WBN                         User's Guide for               0-TI-12.04 Unit 1 & 2              Abnormal and Emergency                Rev. 0000 Operating Instructions              Page 7 of 57 2.1.2   Mode Applicability of the EOIs The EOIs are written to mitigate emergency transients initiated when the unit is at hot or power conditions.
A. The guidance for operator action in the EOIs assumes that the safety-related equipment required by Tech Specs in Mode 1 or Mode 2 is available for use.
B. The operating crew should implement the EOI network whenever reactor trip or safety injection events are initiated with the unit in Modes 1, 2, or 3.
C. The operating crew should implement the EOI network for the complete loss of shutdown power event with the unit in Modes 1, 2, 3, or 4.
D. Implementation of the EOI network in Mode 4 requires the operating crew to consider plant conditions and each specific instructions applicability.
The EOIs are
: 1. The EOI network assumes that the Residual Heat Removal (RHR) system never REQUIRED is aligned for its Emergency Core Cooling mode.
to be used in MODE 5.                2. Although most of the FRs can be utilized to respond to events during Mode 4 conditions, they assume ECCS equipment has operated and steam generators are available and required for heat removal.
: 3. Events (other than complete loss of shutdown power) initiated with the unit in Modes 4, 5, or 6 should be mitigated by implementation of the Abnormal Operating Instructions (AOIs).
: 4. The operating team should consider implementation of the EOI network if events initiated in Modes 4 or 5 result in plant heat-up to Mode 3.
: 5. A specific task (e.g., alignment of RHR sump recirc to SI pump suction) that is detailed in the EOIs may be appropriate during an unanticipated event. When considering such actions, the crew must be cautious and NOT apply the instruction out of context.
 
Frequency 1.4 1.4 Frequency EXAMPLES         EXAMPLE 1.4-2 (continued)
SURVEILLANCE REQUIREMENTS SURVEILLANCE                               FREQUENCY Verify flow is within limits.                               Once within 12 hours after 25% RTP AND 24 hours thereafter The first performance of a Example 1.4-2 has two Frequencies. The first is a one time performance T/S required SR              Frequency, and the second is of the type shown in Example 1.4-1. The logical (with a thereafter          connector "AND" indicates that both Frequency requirements must be met. Each frequency) does              time reactor power is increased from a power level < 25% RTP to 25% RTP, not receive the              the Surveillance must be performed within 12 hours.
alleviation of SR            The use of "once" indicates a single performance will satisfy the specified 3.0.2.                      Frequency (assuming no other Frequencies are connected by "AND"). This type of Frequency does not qualify for the extension allowed by SR 3.0.2. "Thereafter" indicates future performances must be established per SR 3.0.2, but only after a specified condition is first met (i.e., the "once" performance in this example). If reactor power decreases to < 25% RTP, the measurement of both intervals stops. New intervals start upon reactor power reaching 25% RTP.
(continued)
Watts Bar-Unit 1                               1.4-3


Exceptions to this Specification are stated in the individual Specifications. SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed. If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.
SR Applicability 3.0 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1          SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3.
Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.
SR 3.0.2          The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.
For Frequencies specified as "once," the above interval extension does not apply.
If a Completion Time requires periodic performance on a "once per . . ." basis, the above Frequency extension applies to each performance after the initial performance.
Exceptions to this Specification are stated in the individual Specifications.
SR 3.0.3         If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed.
If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.
(continued)
Watts Bar-Unit 1                              3.0-4                                    Amendment 42


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 3 of 16  II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]: A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)] Some examples of SRO exam items for this topic include:
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]:
A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)]
Some examples of SRO exam items for this topic include:
* Reporting requirements when the maximum licensed thermal power output is exceeded.
* Reporting requirements when the maximum licensed thermal power output is exceeded.
* Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems, fire doors, etc.
* Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems, fire doors, etc.
* The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
* The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
* National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
* National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
* Processes for TS and FSAR changes. Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic. B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)] Some examples of SRO exam items for this topic include:
* Processes for TS and FSAR changes.
Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic.
B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]
Some examples of SRO exam items for this topic include:
* Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
* Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
* Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
* Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
* Knowledge of TS bases that are required to analyze TS required actions and terminology.
* Knowledge of TS bases that are required to analyze TS required actions and terminology.
* Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).
* Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of  1 hour action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.  
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of  1 hour action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. ROs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Page 3 of 16


SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. RO's are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 4 of 16 ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)         Above this line RO knowledge Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 5 of 16 Figure 1:  Screening for SRO-only linked to 10 CFR 55.43(b)(2) (Tech Specs)       
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)
RO knowledge Above this line Page 4 of 16


Can question be answered solely by knowing  1 hour TS/TRM Action? RO question YesNoCan question be answered solely by knowing the LCO/TRM information listed "above-the-line?" YesRO question NoCan question be answered solely by knowing the TS Safety Limits? YesRO question NoDoes the question involve one or more of the following for TS, TRM, or ODCM?
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)
(Tech Specs)
Can question be answered solely by knowing  1       Yes hour TS/TRM Action?                                               RO question No Can question be answered solely by knowing the       Yes LCO/TRM information listed above-the-line?                       RO question No Can question be answered solely by knowing the       Yes TS Safety Limits?                                                 RO question No Does the question involve one or more of the following for TS, TRM, or ODCM?
* Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
* Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)
* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)                                 Yes      SRO-only
* Knowledge of TS bases that is required to analyze TS required actions and terminology SRO-only question YesNoQuestion might not be linked to 10 CFR 55.43(b)(2) for SRO-only Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 6 of 16  C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)] Some examples of SRO exam items for this topic include:
* Knowledge of TS bases that is required to analyze TS                       question required actions and terminology The question No requires the knowledge of SR 3.0.2.
Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only Page 5 of 16
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]
Some examples of SRO exam items for this topic include:
* 10 CFR 50.59 screening and evaluation processes.
* 10 CFR 50.59 screening and evaluation processes.
* Administrative processes for temporary modifications.
* Administrative processes for temporary modifications.
* Administrative processes for disabling annunciators.
* Administrative processes for disabling annunciators.
* Administrative processes for the installation of temporary instrumentation.
* Administrative processes for the installation of temporary instrumentation.
* Processes for changing the plant or plant procedures. Section IV provides an example of a satisfactory SRO-only question related to this topic. D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.       [10 CFR 55.43(b)(4)] Some examples of SRO exam items for this topic include:
* Processes for changing the plant or plant procedures.
Section IV provides an example of a satisfactory SRO-only question related to this topic.
D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
[10 CFR 55.43(b)(4)]
Some examples of SRO exam items for this topic include:
* Process for gaseous/liquid release approvals, i.e., release permits.
* Process for gaseous/liquid release approvals, i.e., release permits.
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures. Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits. SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc. E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedure's overall mitigative strategy or purpose.  
Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.
 
SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.
The applicant's knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedure's content is required to correctly answer the written test item, for example:
E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.
The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:
Page 6 of 16


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 7 of 16
The question Clarification Guidance for SRO-only Questions requires the Rev 1 (03/11/2010) knowledge of the
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
* Knowledge of when to implement attachments and appendices, includingMODES of how to coordinate these items with procedure steps.               applicability for the
* Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
* Knowledge of diagnostic steps and decision points in the emergency EOP set (i.e. the operating procedures (EOP) that involve transitions to event specific sub-knowledge    of procedures or emergency contingency procedures.                   which modes
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
* Knowledge of administrative procedures that specify hierarchy, require the use of implementation, and/or coordination of plant normal, abnormal, and the EOPs and emergency procedures.                                              which require the use of the AOIs).
SRO-only knowledge should not be claimed for questions that can be answered solely using "systems knowledge"; e.g.:
SRO-only knowledge should not be claimed for questions that can be answered solely using systems knowledge; e.g.:
* how the system works.
* how the system works.
* system flow path.
* system flow path.
* component locations, etc. SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:
* component locations, etc.
SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:
* the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
* the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
* any AOP entry condition.
* any AOP entry condition.
* plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
* plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
* immediate operator actions of a procedure. Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.        
* immediate operator actions of a procedure.
Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.
Page 7 of 16


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 8 of 16  Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5) (Assessment and selection of procedures)      
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
 
Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)
Can the question be answered solely by knowing "systems knowledge", i.e., how the system works, flowpath, logic, component location? RO question YesNoCan the question be answered solely by knowing immediate operator actions? YesCan the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?YesNoDoes the question require one or more of the following?
(Assessment and selection of procedures)
Can the question be answered solely by knowing systems knowledge, i.e., how the system works,       Yes RO question flowpath, logic, component location?
No Can the question be answered solely by knowing immediate operator actions?                           Yes    RO question No Can the question be answered solely by knowing entry conditions for AOPs or plant parameters         Yes RO question that require direct entry to major EOPs?
No Can the question be answered solely by knowing the purpose, overall sequence of events, or            Yes RO question overall mitigative strategy of a procedure?
No Does the question require one or more of the following?
* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps                                           Yes        SRO-only
* Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
* Knowledge of diagnostic steps and decision points in the             question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?NoNoYesQuestion might not be linked to 10 CFR 55.43(b)(5) for SRO-only NoRO question RO question RO question 
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16
: 85. Given the following conditions: - The MAXIMUM observed Containment pressure was 0.1 psig. - An overpressure condition exists on the #1 SG.
- The operating crew has completed steps 1 and 2 of FR-H.2.
- The crew CHECKS affected SG NR level and notes the following:  1-M-4 Indications    ICS indication  In accordance with FR-H.2 and the associated basis document, which ONE of the following identifies the crew response? The crew will identify that __________. A. neither level shown above indicates that the SG is potentially filled solid with water and as such, the crew will REMAIN in FR-H.2 B. ONLY the level shown on ICS indicates that the SG is potentially filled solid with water but because ONLY the 1-M-4 indications may used for a decision point, the crew will REMAIN in FR-H.2 C. BOTH the levels shown on ICS and 1-M-4 indicate that the SG is potentially filled solid with water; the crew will TRANSITION to FR-H.3 D. ONLY the level shown on ICS indicates that the SG is potentially filled solid with water and because the ICS data may be used for a decision point, the crew will TRANSITION to FR-H.3  NOTE: FR-H.2, Steam Generator Overpressure FR-H.3, Steam Generator High Level CORRECT ANSWER:B DISTRACTOR ANALYSIS:  A. Incorrect: As seen in the Westinghouse basis document for FR-H.2, Steam Generator Overpressure, "If the level is greater than [93%], the SG water level may be above the narrow range or the SG may even be filled solid with water."  Therefore, the value observed on ICS (the computer screen) is greater than 93% and as such represents that the SG may be filled with water. Therefore, it is incorrect to believe that neither level shown indicates that the S/G is potentially filled solid. B. Correct: As described, the 94% NR SG level observed on ICS indicates that the SG is potentially filled solid with water. As seen in TI-12.04, "User's Guide for Abnormal and Emergency Operating Instructions," "During performance of the EOI set, the operator is required to utilize PAM instruments when they are provided on the control board."  Therefore, as the 1-M-4 indications are all 92% and therefore, less than the value requiring a transition to FR-H.3, the crew will remain in FR-H.2. C. Incorrect: Again, only the ICS value is greater than 93%; therefore, it is incorrect to believe that all indications relate that the SG is potentially filled solid with water. It would be correct that a transition to FR-H.3 would be required if the PAM grade instruments were in excess of the setpoint; however, in this case they are not and as such the crew will remain in FR-H.2. D. Incorrect: While it is true that the ICS data suggests that the SG is potentially filled solid, it is not true (as previously discussed) that the data can be used to make a transition to FR-H.3.
Question Number: 85  Tier:  1 Group:  2  K/A: WE13  Steam Generator Over-pressure G2.4.3 Ability to identify post-accident instrumentation Importance Rating: 3.7  3.9


10 CFR Part 55: (CFR: 41.6 / 45.4)
85.
Given the following conditions:
    - The MAXIMUM observed Containment pressure was 0.1 psig.
    - An overpressure condition exists on the #1 SG.
    - The operating crew has completed steps 1 and 2 of FR-H.2.
    - The crew CHECKS affected SG NR level and notes the following:
1-M-4 Indications                                ICS indication In accordance with FR-H.2 and the associated basis document, which ONE of the following identifies the crew response?
The crew will identify that __________.
NOTE: FR-H.2, Steam Generator Overpressure FR-H.3, Steam Generator High Level A. neither level shown above indicates that the SG is potentially filled solid with water and as such, the crew will REMAIN in FR-H.2 B. ONLY the level shown on ICS indicates that the SG is potentially filled solid with water but because ONLY the 1-M-4 indications may used for a decision point, the crew will REMAIN in FR-H.2 C.      BOTH the levels shown on ICS and 1-M-4 indicate that the SG is potentially filled solid with water; the crew will TRANSITION to FR-H.3 D.      ONLY the level shown on ICS indicates that the SG is potentially filled solid with water and because the ICS data may be used for a decision point, the crew will TRANSITION to FR-H.3


10CFR55.43.b: 10 CFR 55.43(b)(5)  
CORRECT ANSWER:                                                                B DISTRACTOR ANALYSIS:
A. Incorrect: As seen in the Westinghouse basis document for FR-H.2, Steam Generator Overpressure, If the level is greater than [93%], the SG water level may be above the narrow range or the SG may even be filled solid with water. Therefore, the value observed on ICS (the computer screen) is greater than 93% and as such represents that the SG may be filled with water. Therefore, it is incorrect to believe that neither level shown indicates that the S/G is potentially filled solid.
B. Correct: As described, the 94% NR SG level observed on ICS indicates that the SG is potentially filled solid with water. As seen in TI-12.04, User's Guide for Abnormal and Emergency Operating Instructions, During performance of the EOI set, the operator is required to utilize PAM instruments when they are provided on the control board.
Therefore, as the 1-M-4 indications are all 92% and therefore, less than the value requiring a transition to FR-H.3, the crew will remain in FR-H.2.
C. Incorrect: Again, only the ICS value is greater than 93%; therefore, it is incorrect to believe that all indications relate that the SG is potentially filled solid with water. It would be correct that a transition to FR-H.3 would be required if the PAM grade instruments were in excess of the setpoint; however, in this case they are not and as such the crew will remain in FR-H.2.
D. Incorrect: While it is true that the ICS data suggests that the SG is potentially filled solid, it is not true (as previously discussed) that the data can be used to make a transition to FR-H.3.


K/A Match: K/A is matched because the applicant is required to correctly implement the steps of FR-H.2 given displays of both PAM instrumentation and regular instrumentation.
Question Number:      85 Tier:    1  Group:      2 K/A:   WE13 Steam Generator Over-pressure G2.4.3 Ability to identify post-accident instrumentation Importance Rating:    3.7 3.9 10 CFR Part 55:    (CFR: 41.6 / 45.4) 10CFR55.43.b:      10 CFR 55.43(b)(5)
K/A          K/A is matched because the applicant is required to correctly Match:      implement the steps of FR-H.2 given displays of both PAM instrumentation and regular instrumentation.
Technical  
Technical  


==Reference:==
==Reference:==
FR-H.2, Steam Generator Overpressure FR-H.3, Steam Generator High Level TI-12.04, User's Guide for Abnormal and Emergency Operating Instructions Proposed references to be provided: None  Learning Objective: 3-OT-FRH0001 3. Explain the purpose for and the basis of each step in FR-H procedures 6. Given a set of plant conditions use the FR-H procedures to correctly identify and required procedure transition Cognitive Level:     Higher X Lower     Question Source:     New X Modified Bank   Bank     Question History: New question for the 2015-301 NRC SRO Exam Comments: See the marked up Clarification Guidance for SRO-only Questions.  
FR-H.2, Steam Generator Overpressure FR-H.3, Steam Generator High Level TI-12.04, Users Guide for Abnormal and Emergency Operating Instructions Proposed references       None to be provided:
Learning Objective:       3-OT-FRH0001
: 3. Explain the purpose for and the basis of each step in FR-H procedures
: 6. Given a set of plant conditions use the FR-H procedures to correctly identify and required procedure transition Cognitive Level:
Higher             X Lower Question Source:
New                 X Modified Bank Bank Question History:         New question for the 2015-301 NRC SRO Exam Comments:                 See the marked up Clarification Guidance for SRO-only Questions.
 
STEP DESCRIPTION TABLE FOR FR-H.2                Step 3__
STEP:      Check Affected SG(s) Narrow Range Level - LESS THAN (M.08)%
[(M.09)% FOR ADVERSE CONTAINMENT]
93% for WBN PURPOSE:    To determine if overfilling the affected SG is a potential cause of the overpressurization BASIS:
The operator should check the affected SG level to ensure that it is not above (M.08)% [(M.09)% for adverse containment]. If the level is greater than (M.08)% [(M.09)% for adverse containment], the SG water level may be above the narrow range or the SG may even be filled solid with water. For this case, the operator is transferred to FR-H.3, RESPONSE TO STEAM GENERATOR HIGH LEVEL, to address the high water level condition.                    At 94%, the S/G is potentially filled ACTIONS:                                                      solid with water.
o Determine if the affected SG narrow range level is less than (M.08)%
[(M.09)% for adverse containment]
o Transfer to FR-H.3, RESPONSE TO STEAM GENERATOR HIGH LEVEL, Step 1 INSTRUMENTATION:
SG narrow range level CONTROL/EQUIPMENT:
N/A KNOWLEDGE:
N/A PLANT-SPECIFIC INFORMATION:
o (M.08) SG level at the upper tap, including allowances for normal channel accuracy.
o (M.09) SG level at the upper tap, including allowances for normal channel accuracy, post-accident transmitter errors, and reference leg process errors.
FR-H.2 Background                    11                HP-Rev. 2, 4/30/2005 HFRH2BG.doc
 
WBN                        User's Guide for                  0-TI-12.04 Unit 1 & 2              Abnormal and Emergency                Rev. 0000 Operating Instructions              Page 12 of 57 2.2.1    Cautions and Notes (continued)
C. CAUTIONS and NOTES are introduced by their designator in bold face type.
: 1. The designator is followed by the text extending across the entire page with note text appearing in standard type and caution text appearing in bold face type.
: 2. If multiple cautions or notes are applicable to a step then each caution or note included after the initial designator is distinguished by a preceding bullet.
D. In general, CAUTIONS and NOTES apply to the step which they precede.
E. CAUTIONS and NOTES which precede the first operator action step may also apply throughout the instruction.
F. When CAUTIONS or NOTES are communicated (Read by procedure reader) they are to be communicated through directive communication addressed to individual(s) and verified via 3-way communication.
2.2.2    Use of Instrumentation A. Post Accident Monitoring (PAM) instrumentation is provided on the main control board as determined by design requirements.
: 1. The control room complies with Reg. Guide 1.97 requirements by providing Because PAM S/G                the operator with the required PAM instruments.
NR is provided on the MCR board 1-          2. The control board PAM instruments are uniquely labeled to identify them M-4, it is required            as PAM instrumentation.
to be used.
: a. Most PAM instruments have black background instrumentation labels.
: b. Other PAM instruments are identified by a small box located on the instrument label with the designator C1 or C2 inside the box.
: 3. During performance of the EOI set, the operator is required to utilize PAM instruments when they are provided on the control board.
: 4. The operator should compare redundant instruments when they are provided.
: 5. Some parameters evaluated during performance of the EOIs do NOT have PAM grade instrumentation provided on the control board.
: a. The operator should monitor these parameters with available instrumentation.


STEPDESCRIPTIONTABLEFORFR-H.2Step3__STEP:CheckAffectedSG(s)NarrowRangeLevel-LESSTHAN(M.08)%[(M.09)%FORADVERSECONTAINMENT]PURPOSE:TodetermineifoverfillingtheaffectedSGisapotentialcauseoftheoverpressurizationBASIS:TheoperatorshouldchecktheaffectedSGleveltoensurethatitisnotabove(M.08)%[(M.09)%foradversecontainment].Ifthelevelisgreaterthan(M.08)%[(M.09)%foradversecontainment],theSGwaterlevelmaybeabovethenarrowrangeortheSGmayevenbefilledsolidwithwater.Forthiscase,theoperatoristransferredtoFR-H.3,RESPONSETOSTEAMGENERATORHIGHLEVEL,toaddressthehighwaterlevelcondition.ACTIONS:oDetermineiftheaffectedSGnarrowrangelevelislessthan(M.08)%[(M.09)%foradversecontainment]oTransfertoFR-H.3,RESPONSETOSTEAMGENERATORHIGHLEVEL,Step1INSTRUMENTATION:SGnarrowrangelevelCONTROL/EQUIPMENT:N/AKNOWLEDGE:N/APLANT-SPECIFICINFORMATION:o(M.08)SGlevelattheuppertap,includingallowancesfornormalchannelaccuracy.o(M.09)SGlevelattheuppertap,includingallowancesfornormalchannelaccuracy,post-accidenttransmittererrors,andreferencelegprocesserrors.FR-H.2BackgroundHFRH2BG.doc11HP-Rev.2,4/30/2005 WBN Unit 1 & 2 User's Guide for  Abnormal and Emergency  Operating Instructions 0-TI-12.04 Rev. 0000 Page 12 of 57  2.2.1 Cautions and Notes (continued)    C. CAUTIONS and NOTES are introduced by their designator in bold face type. 1. The designator is followed by the text extending across the entire page with note text appearing in standard type and caution text appearing in bold face type. 2. If multiple cautions or notes are applicable to a step then each caution or note included after the initial designator is distinguished by a preceding bullet. D. In general, CAUTIONS and NOTES apply to the step which they precede. E. CAUTIONS and NOTES which precede the first operator action step may also apply throughout the instruction. F. When CAUTIONS or NOTES are communicated (Read by procedure reader) they are to be communicated through directive communication addressed to individual(s) and verified via 3-way communication. 2.2.2 Use of Instrumentation  A. Post Accident Monitoring (PAM) instrumentation is provided on the main control board as determined by design requirements. 1. The control room complies with Reg. Guide 1.97 requirements by providing the operator with the required PAM instruments. 2. The control board PAM instruments are uniquely labeled to identify them as PAM instrumentation. a. Most PAM instruments have black background instrumentation labels. b. Other PAM instruments are identified by a small box located on the instrument label with the designator "C1" or "C2" inside the box. 3. During performance of the EOI set, the operator is required to utilize PAM instruments when they are provided on the control board. 4. The operator should compare redundant instruments when they are provided. 5. Some parameters evaluated during performance of the EOIs do NOT have PAM grade instrumentation provided on the control board. a. The operator should monitor these parameters with available instrumentation.
WBN               Steam Generator Overpressure         FR-H.2 Unit 1                                                  Rev. 0006 Step   Action/Expected Response                 Response Not Obtained 3.0     OPERATOR ACTIONS
WBN Unit 1 Steam Generator Overpressure FR-H.2 Rev. 0006    Page 3 of 6  Step Action/Expected Response Response Not Obtained 3.0 OPERATOR ACTIONS 1. IDENTIFY affected S/G(s): a. Any S/G pressure greater than or equal to 1220 psig. a. IF press in all S/Gs less than 1220 psig, THEN RETURN TO Instruction in effect. 2. ENSURE MFW isolated to affected S/Gs:
: 1.     IDENTIFY affected S/G(s):
: a. Any S/G pressure greater             a. IF press in all S/Gs less than than or equal to 1220 psig.              1220 psig, THEN RETURN TO Instruction in effect.
: 2.     ENSURE MFW isolated to                   Manually CLOSE valves, AND affected S/Gs:
STOP pumps, as necessary.
* S/G MFW isolation and bypass isolation valves CLOSED.
* S/G MFW isolation and bypass isolation valves CLOSED.
* S/G MFW reg and bypass reg valves CLOSED.
* S/G MFW reg and bypass reg           IF valves can NOT be closed, THEN valves CLOSED.
CLOSE #1 heater outlet valves.
* MFP A and B TRIPPED.
* MFP A and B TRIPPED.
* Standby MFP STOPPED.
This is the
* Cond demin pumps TRIPPED.
* Standby MFP STOPPED.                                       transition being
* Cond booster pumps TRIPPED. Manually CLOSE valves, AND STOP pumps, as necessary.
* Cond demin pumps TRIPPED.                                  tested.
IF valves can NOT be closed, THEN  CLOSE #1 heater outlet valves. 3. CHECK affected S/Gs NR level less than 93% [85% ADV]. ** GO TO FR-H.3, STEAM GENERATOR HIGH LEVEL. 4. DEPRESSURIZE affected S/Gs:
* Cond booster pumps TRIPPED.
* OPEN S/G PORVs, OR
: 3.     CHECK affected S/Gs NR level             ** GO TO FR-H.3, STEAM less than 93% [85% ADV].                  GENERATOR HIGH LEVEL.
* OPEN S/G MSIV bypass valves, OR
The adverse
* OPEN S/G steam supply to TD AFW pump, OR
: 4.     DEPRESSURIZE affected S/Gs:
setpoint is not in
* OPEN S/G PORVs,                                     effect for this question.
OR
* OPEN S/G MSIV bypass valves,         Because the PAM instruments exist on 1-M-4 for SG NR level, they must be OR                                   used. The transition to FR-H.3 cannot be made using the ICS value. It must
* OPEN S/G steam supply be made using the PAM instruments' to TD AFW pump, readings. This is seen in TI-12.04.
OR
* OPEN S/G blowdown valves.
* OPEN S/G blowdown valves.
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 6 of 16  C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)] Some examples of SRO exam items for this topic include:
Page 3 of 6
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]
Some examples of SRO exam items for this topic include:
* 10 CFR 50.59 screening and evaluation processes.
* 10 CFR 50.59 screening and evaluation processes.
* Administrative processes for temporary modifications.
* Administrative processes for temporary modifications.
* Administrative processes for disabling annunciators.
* Administrative processes for disabling annunciators.
* Administrative processes for the installation of temporary instrumentation.
* Administrative processes for the installation of temporary instrumentation.
* Processes for changing the plant or plant procedures. Section IV provides an example of a satisfactory SRO-only question related to this topic. D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.       [10 CFR 55.43(b)(4)] Some examples of SRO exam items for this topic include:
* Processes for changing the plant or plant procedures.
Section IV provides an example of a satisfactory SRO-only question related to this topic.
D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
[10 CFR 55.43(b)(4)]
Some examples of SRO exam items for this topic include:
* Process for gaseous/liquid release approvals, i.e., release permits.
* Process for gaseous/liquid release approvals, i.e., release permits.
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures. Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits. SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc. E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedure's overall mitigative strategy or purpose.  
Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.
 
SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.
The applicant's knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedure's content is required to correctly answer the written test item, for example:
E. Assessment of facility conditions and selection of appropriate procedures The applicant must    during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
assess the condition of the SG    This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or by using both PAM section of a procedure to mitigate, recover, or with which to proceed. One and regular area of SRO level knowledge (with respect to selecting a procedure) is instrumentation knowledge of the content of the procedure versus knowledge of the and then              procedures overall mitigative strategy or purpose.                   Detailed implement the                                                                                knowledge of the correct procedure. The applicants knowledge can be evaluated at the level of 10 CFR transitional step is 55.43(b)(5) by ensuring that the additional knowledge of the procedures required (i.e.
content is required to correctly answer the written test item, for example:
knowledge of the setpoint).
Page 6 of 16


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 7 of 16
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
* Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
* Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
SRO-only knowledge should not be claimed for questions that can be answered solely using "systems knowledge"; e.g.:
SRO-only knowledge should not be claimed for questions that can be answered solely using systems knowledge; e.g.:
* how the system works.
* how the system works.                                           The procedures
* system flow path.
* system flow path.                                               FR-H.2 and FR-H.3
* component locations, etc. SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:
* component locations, etc.                                       are both yellow path functional SRO-only knowledge should not be claimed for questions that can restorations.
be answered solely using fundamental knowledge of:
* the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
* the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
* any AOP entry condition.
* any AOP entry condition.
* plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
* plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
* immediate operator actions of a procedure. Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.        
* immediate operator actions of a procedure.
 
Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 8 of 16   Figure 2:  Screening for SRO-only linked to 10 CFR 55.43(b)(5) (Assessment and selection of procedures)       
Page 7 of 16


Can the question be answered solely by knowing "systems knowledge", i.e., how the system works, flowpath, logic, component location? RO question YesNoCan the question be answered solely by knowing immediate operator actions? YesCan the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?YesNoDoes the question require one or more of the following?
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)
(Assessment and selection of procedures)
Can the question be answered solely by knowing systems knowledge, i.e., how the system works,       Yes RO question flowpath, logic, component location?
No Can the question be answered solely by knowing immediate operator actions?                           Yes    RO question No Can the question be answered solely by knowing entry conditions for AOPs or plant parameters         Yes RO question that require direct entry to major EOPs?
No Can the question be answered solely by knowing the purpose, overall sequence of events, or            Yes RO question overall mitigative strategy of a procedure?
No Does the question require one or more of the following?
* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps                                           Yes        SRO-only
* Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
* Knowledge of diagnostic steps and decision points in the             question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?NoNoYesQuestion might not be linked to 10 CFR 55.43(b)(5) for SRO-only NoRO question RO question RO question 
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16
: 86. Given the following timeline:  00:00:00- Unit 1 is at 100%. 00:0 1:00 - The primary +15VDC power supply trips in train B SSPS.
- The secondary +15VDC power supply picked up the SSPS logic circuit card load. - SSPS-B GEN WARNING (115-A) is LIT. 06:00:00- A troubleshooting/repair work order is authorized and placed in WORKING status. 08:00:00- The SSPS engineer wishes to block the General Warning signal from the failed +15VDC power supply to facilitate troubleshooting.  - He proposes performing the following under the work order authorized at 06:00:00: o Lift the output wires for the failed +15VDC power supply. o Install a jumper to supply the failed input to the general warning circuit from the functional +15VDC power supply. o The jumper is expected to be installed for two weeks.
Which ONE of the following describes the operability of the B Train SSPS AND whether or not a Technical Evaluation (TE) review is required?


In accordance with T/S LCO 3.3.1, RTS Instrumentation B Train SSPS is ____(1)____ at 00:01:01. Performance of the recommendations made at 08:00:00 ____(2)____ require a TE review In accordance with NPG-SPP-09.5, Temporary Modifications. REFERENCE PROVIDED A. (1) operable (2) WILL B. (1) operable (2) WILL NOT C. (1) inoperable (2) WILL D. (1) inoperable (2) WILL NOT CORRECT ANSWER:A DISTRACTOR ANALYSIS:  A. Correct: The Westinghouse manual for the SSPS system, WBN-VTD-W120-2454, indicates that two 15VDC power supplies exist within the logic bay of each SSPS train. These supplies feed redundantly to the 15VDC buses; they do this through the action of an auctioneering circuit. This design promotes the continuity of power of the generating unit should a solitary power supply be lost. Because of this design, there is no impact to the operability of SSPS given the loss of a single low voltage power supply. Therefore, it is correct that the B SSPS train was operable immediately following the loss of one of the two +15VDC power supplies.
86.
It is correct that a TE be required to implement the desires of the system engineer. This is seen in section 3.7 E which states:  A Technical Evaluation (TE) review is required for all WO-TMs. B. Incorrect: While it is correct that the B SSPS train remains operable following the loss of a solitary +15VDC power supply, it is not correct that a Technical Evaluation not be required. It is plausible (for many reasons) why this evaluation would not be required. Firstly, if the applicant may believe that the exclusion 2.2 I.2 of the aforementioned SPP applied. This exclusion states that Connections to permanently installed test jacks to take a reading are excluded from the restrictions of the procedure. The proposed connections for the jumper to be installed are test jacks and are normally used to measure the output of the two 15VDC power supplies. However, his proposal is not to conduct measurements but to cross connect power supplies; therefore, the exclusion is not applicable. Next, (as seen in 2.2 K), Temporary changes that are continuously attended are excluded. However, the stem of the question details that the jumper will not be attended and as such is not allowed the liberty of the described exclusion. C. Incorrect: While it is correct that the jumper does require that a technical evaluation review be conducted, it is not correct that the train of SSPS is rendered inoperable following the failure of a +15VDC power supply. D. Incorrect: As discussed it is incorrect and yet plausible that the SSPS train is rendered inoperable upon the loss of one of the two +15VDC power supplies. Also, it is incorrect and yet plausible that a Technical Evaluation would not be required.
Given the following timeline:
Question Number: 86  Tier:  2 Group:  1 K/A: 012 Reactor Protection System 2.2 Equipment Control 2.2.36 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.
00:00:00 - Unit 1 is at 100%.
Importance Rating: 3.1  4.2
00:0 1:00 - The primary +15VDC power supply trips in train B SSPS.
                -  The secondary +15VDC power supply picked up the SSPS logic circuit card load.
                -  SSPS-B GEN WARNING (115-A) is LIT.
06:00:00 -      A troubleshooting/repair work order is authorized and placed in WORKING status.
08:00:00 -      The SSPS engineer wishes to block the General Warning signal from the failed +15VDC power supply to facilitate troubleshooting.
                -  He proposes performing the following under the work order authorized at 06:00:00:
o Lift the output wires for the failed +15VDC power supply.
o Install a jumper to supply the failed input to the general warning circuit from the functional +15VDC power supply.
o The jumper is expected to be installed for two weeks.
Which ONE of the following describes the operability of the B Train SSPS AND whether or not a Technical Evaluation (TE) review is required?
In accordance with T/S LCO 3.3.1, RTS Instrumentation B Train SSPS is ____(1)____ at 00:0 1 :0 1.
Performance of the recommendations made at 08:00:00 ____(2)____ require a TE review In accordance with NPG-SPP-09.5, Temporary Modifications.
REFERENCE PROVIDED A.       (1) operable (2) WILL B.       (1) operable (2) WILL NOT C.       (1) inoperable (2) WILL D.       (1) inoperable (2) WILL NOT


10 CFR Part 55: (CFR: 41.10 / 43.2 / 45.13) 10CFR55.43.b: 10 CFR 55.43(b)(2) and 10 CFR 55.43(b)(3)
CORRECT ANSWER:                                                               A DISTRACTOR ANALYSIS:
A. Correct: The Westinghouse manual for the SSPS system, WBN-VTD-W120-2454, indicates that two 15VDC power supplies exist within the logic bay of each SSPS train. These supplies feed redundantly to the 15VDC buses; they do this through the action of an auctioneering circuit. This design promotes the continuity of power of the generating unit should a solitary power supply be lost. Because of this design, there is no impact to the operability of SSPS given the loss of a single low voltage power supply. Therefore, it is correct that the B SSPS train was operable immediately following the loss of one of the two
              +15VDC power supplies.
It is correct that a TE be required to implement the desires of the system engineer. This is seen in section 3.7 E which states: A Technical Evaluation (TE) review is required for all WO-TMs.
B. Incorrect: While it is correct that the B SSPS train remains operable following the loss of a solitary +15VDC power supply, it is not correct that a Technical Evaluation not be required. It is plausible (for many reasons) why this evaluation would not be required. Firstly, if the applicant may believe that the exclusion 2.2 I.2 of the aforementioned SPP applied. This exclusion states that Connections to permanently installed test jacks to take a reading are excluded from the restrictions of the procedure. The proposed connections for the jumper to be installed are test jacks and are normally used to measure the output of the two 15VDC power supplies. However, his proposal is not to conduct measurements but to cross connect power supplies; therefore, the exclusion is not applicable. Next, (as seen in 2.2 K),
Temporary changes that are continuously attended are excluded.
However, the stem of the question details that the jumper will not be attended and as such is not allowed the liberty of the described exclusion.
C. Incorrect: While it is correct that the jumper does require that a technical evaluation review be conducted, it is not correct that the train of SSPS is rendered inoperable following the failure of a +15VDC power supply.
D. Incorrect: As discussed it is incorrect and yet plausible that the SSPS train is rendered inoperable upon the loss of one of the two +15VDC power supplies. Also, it is incorrect and yet plausible that a Technical Evaluation would not be required.


K/A Match: K/A is matched because the applicant is required to analyze the effect that a degraded 15VDC power supply has on a SSPS train using both systems knowledge and the contents of the T/S bases. Technical  
Question Number:        86 Tier:    2  Group:      1 K/A:    012 Reactor Protection System 2.2 Equipment Control 2.2.36 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.
Importance Rating:      3.1 4.2 10 CFR Part 55:      (CFR: 41.10 / 43.2 / 45.13) 10CFR55.43.b:        10 CFR 55.43(b)(2) and 10 CFR 55.43(b)(3)
K/A Match:   K/A is matched because the applicant is required to analyze the effect that a degraded 15VDC power supply has on a SSPS train using both systems knowledge and the contents of the T/S bases.
Technical  


==Reference:==
==Reference:==
PER Vault Summary for PER 3516 Bases for T/S LCO 3.3.1 NPG-SPP-09.5, Temporary Modifications Westinghouse SSPS Technical Manual, WBN-VTD-W120-2454 50.59 screen for WO 01-008855-000 Proposed references to be provided: None  Learning Objective:   Cognitive Level:     Higher X Lower     Question Source:     New X Modified Bank   Bank     Question History: New question for the 2015-301 NRC SRO Exam Comments: See the marked up Clarification Guidance for SRO-only Questions.  
PER Vault Summary for PER 3516 Bases for T/S LCO 3.3.1 NPG-SPP-09.5, Temporary Modifications Westinghouse SSPS Technical Manual, WBN-VTD-W120-2454 50.59 screen for WO 01-008855-000 Proposed references to     None be provided:
Learning Objective:
Cognitive Level:
Higher               X Lower Question Source:
New                 X Modified Bank Bank Question History:           New question for the 2015-301 NRC SRO Exam Comments:                   See the marked up Clarification Guidance for SRO-only Questions.


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 3 of 16  II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]: A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)] Some examples of SRO exam items for this topic include:
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]:
A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)]
Some examples of SRO exam items for this topic include:
* Reporting requirements when the maximum licensed thermal power output is exceeded.
* Reporting requirements when the maximum licensed thermal power output is exceeded.
* Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems, fire doors, etc.
* Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems, fire doors, etc.
* The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
* The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
* National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
* National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
* Processes for TS and FSAR changes. Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic. B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)] Some examples of SRO exam items for this topic include:
* Processes for TS and FSAR changes.
* Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic.
* Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]
* Knowledge of TS bases that are required to analyze TS required actions and terminology.
The question        Some examples of SRO exam items for this topic include:
requires the applicant to
* Application of Required Actions (Section 3) and Surveillance determine whether        Requirements (SR) (Section 4) in accordance with rules of application or not a train of        requirements (Section 1).
SSPS remains
* Application of generic Limiting Condition for Operation (LCO) operable following        requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
a loss of a power
* Knowledge g of TS bases that are required to analyze TS required actions supply.                  and terminology.
* Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).
* Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of1 hour action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.  
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of  1 hour action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. ROs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Page 3 of 16


SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. ROs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the B. Facility operating limitations in the TS and their bases.[10 CFR 55.43(b)(2)] *Knowledge of TS bases that are required to analyze TS required actions gand terminology. Thequestionrequiresthe applicantto determinewhether ornotatrainof SSPSremains operablefollowing alossofapower supply.
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 4 of 16 ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)         Above this line RO knowledge Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 5 of 16 Figure 1:  Screening for SRO-only linked to 10 CFR 55.43(b)(2) (Tech Specs)       
ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)
RO knowledge Above this line Page 4 of 16


Can question be answered solely by knowing hour TS/TRM Action? RO question YesNoCan question be answered solely by knowing the LCO/TRM information listed above-the-line? YesRO question NoCan question be answered solely by knowing the TS Safety Limits? YesRO question NoDoes the question involve one or more of the following for TS, TRM, or ODCM?
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)
(Tech Specs)
Can question be answered solely by knowing  1      Yes hour TS/TRM Action?                                                 RO question No Can question be answered solely by knowing the       Yes LCO/TRM information listed above-the-line?                       RO question No Can question be answered solely by knowing the       Yes TS Safety Limits?                                                   RO question No Does the question involve one or more of the following for TS, TRM, or ODCM?
* Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
* Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)
* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4))                                  Yes      SRO-only
* Knowledge of TS bases that is required to analyze TS required actions and terminology SRO-only question YesNoQuestion might not be linked to 10 CFR 55.43(b)(2) for SRO-only NoNoNo* )Knowledge of TS bases that is required to analyze TSgrequired actions and terminology YesSRO-only questionTherefore,thedeterminationof operabilityofthe trainofSSPSis SROonly.
* Knowledge g of TS bases that is required to analyze TS                     question required actions and terminology No Therefore, the determination of Question might not be linked to operability of the 10 CFR 55.43(b)(2) for SRO-only train of SSPS is SRO only.
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 6 of 16  C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)] Some examples of SRO exam items for this topic include:
Page 5 of 16
* 10 CFR 50.59 screening and evaluation processes.
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
C. Facility y licensee pprocedures required q      to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]
The use of NPG-Some examples of SRO exam items for this topic include:                 SPP-09.5 meets the intent of this
* 10 CFR 50.59 screening   g and evaluation processes.                bullet.
* Administrative processes for temporary modifications.
* Administrative processes for temporary modifications.
* Administrative processes for disabling annunciators.
* Administrative processes for disabling annunciators.
* Administrative processes for the installation of temporary instrumentation.
* Administrative processes for the installation of temporary instrumentation.
* Processes for changing the plant or plant procedures. Section IV provides an example of a satisfactory SRO-only question related to this topic. D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.       [10 CFR 55.43(b)(4)] Some examples of SRO exam items for this topic include:
* Processes for changing the plant or plant procedures.
Section IV provides an example of a satisfactory SRO-only question related to this topic.
D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
[10 CFR 55.43(b)(4)]
Some examples of SRO exam items for this topic include:
* Process for gaseous/liquid release approvals, i.e., release permits.
* Process for gaseous/liquid release approvals, i.e., release permits.
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures. Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits. SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc. E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.  
Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.
SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.
E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.
The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:
Page 6 of 16


The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:
RTS Instrumentation B 3.3.1 Bases APPLICABLE       18. Reactor Trip Breaker Undervoltage and Shunt Trip SAFETY ANALYSES,       Mechanisms (continued)
C. Facility licensee procedures required to obtain authority for design and ypqoperating changes in the facility.[10 CFR 55.43(b)(3)] *gAdministrative processes for temporary modifications.TheuseofNPG-SPP-09.5meets theintentofthis bullet.
LCO, and APPLICABILITY         service. The trip mechanisms are not required to be OPERABLE for trip breakers that are open, racked out, incapable of supplying power to the CRD System, or declared inoperable under Function 17 above.
RTS Instrumentation B 3.3.1 Bases   (continued)       Watts Bar-Unit 1 B 3.3-38      APPLICABLE  18. Reactor Trip Breaker Undervoltage and Shunt Trip SAFETY ANALYSES, Mechanisms (continued) LCO, and APPLICABILITY service. The trip mechanisms are not required to be OPERABLE for trip breakers that are open, racked out, incapable of supplying power to the CRD System, or declared inoperable under Function 17 above. OPERABILITY of both trip mechanisms on each breaker ensures that no single trip mechanism failure will prevent opening any breaker on a valid signal.
OPERABILITY of both trip mechanisms on each breaker ensures that no single trip mechanism failure will prevent opening any breaker on a valid signal.
These trip Functions must be OPERABLE in MODE 1 or 2 when the reactor is critical. In MODE 3, 4, or 5, these RTS trip Functions must be OPERABLE when the RTBs and associated bypass breakers are closed, and the CRD System is capable of rod withdrawal.
These trip Functions must be OPERABLE in MODE 1 or 2 when the reactor is critical. In MODE 3, 4, or 5, these RTS trip Functions must be Note that the SSPS                OPERABLE when the RTBs and associated bypass breakers are closed, remains completely                and the CRD System is capable of rod withdrawal.
: 19. Automatic Trip Logic The LCO requirement for the RTBs (Functions 17 and 18)and Automatic Trip Logic (Function 19) ensures that means are provided to interrupt the power to allow the rods to fall into the reactor core. Each RTB is equipped with an undervoltage coil and a shunt trip coil to trip the breaker open when needed. Each RTB is equipped with a bypass breaker to allow testing of the trip breaker while the unit is at power. The reactor trip signals generated by the RTS Automatic Trip Logic cause the RTBs and associated bypass breakers to open and shut down the reactor.
capable of
: 19. Automatic Trip Logic performing this function following                The LCO requirement for the RTBs (Functions 17 and 18)and Automatic the failure of the                Trip Logic (Function 19) ensures that means are provided to interrupt the single power                      power to allow the rods to fall into the reactor core. Each RTB is equipped with an undervoltage coil and a shunt trip coil to trip the breaker supply.
open when needed. Each RTB is equipped with a bypass breaker to allow testing of the trip breaker while the unit is at power. The reactor trip signals generated by the RTS Automatic Trip Logic cause the RTBs and associated bypass breakers to open and shut down the reactor.
The LCO requires two trains of RTS Automatic Trip Logic to be OPERABLE. Having two OPERABLE channels ensures that random failure of a single logic channel will not prevent reactor trip.
The LCO requires two trains of RTS Automatic Trip Logic to be OPERABLE. Having two OPERABLE channels ensures that random failure of a single logic channel will not prevent reactor trip.
These trip Functions must be OPERABLE in MODE 1 or 2 when the reactor is critical. In MODE 3, 4, or 5, these RTS trip Functions must be OPERABLE when the RTBs and associated bypass breakers are closed, and the CRD System is capable of rod withdrawal.
These trip Functions must be OPERABLE in MODE 1 or 2 when the reactor is critical. In MODE 3, 4, or 5, these RTS trip Functions must be OPERABLE when the RTBs and associated bypass breakers are closed, and the CRD System is capable of rod withdrawal.
The RTS instrumentation satisfies Criterion 3 of the NRC Policy Statement.
The RTS instrumentation satisfies Criterion 3 of the NRC Policy Statement.
(continued)
Watts Bar-Unit 1                          B 3.3-38


Automatic Trip Logic (Function19) ensures that means are provided to interrupt the power to allow the rods to fallinto the reactor core. The LCO requires two trains of RTS Automatic Trip Logic to beOPERABLE. Having two OPERABLE channels ensures that randomfailure of a single logic channel will not prevent reactor trip.NotethattheSSPSremainscompletely capableof performingthis functionfollowing thefailureofthe singlepower supply.
The power supplies are redundant.
Thepowersuppliesareredundant.  


PER Vault Summary Report for PER: 3516PER Number: 3516Status:ARCHIVEStatus By ID: Status Date: Site / Org: WBNUnit: CAP Due Date: Date of Occurence: Long Lead CA Date: Analysis Type: PER Level: PER Summary: Problem Evaluation Report - User InformationOriginator ID: Originator Name: Originator Phone: Originator Email: On Behalf of ID: On Behalf of Name: On Behalf of Phone: On Behalf of Email: PER Details: Problem Evaluation Report - Problem DetailsInitiating Department: Site / Org: WBNSBU: BU: Unit: PER Summary: Asset: Location: Additional Location Details: Plant System: Corrected Immediately: NOImmediate Actions Taken: Reported Date: Date Of Occurence: Asset Site: ONE 15VDC POWER SUPPLY FOR SSPS TR B, LOCATED IN 1-R-50, HAS GONE BAD. THIS BROUGHT IN A GENERAL WARNING FOR B TR SSPS. BTR SSPS IS STILL OPERABLE DUE TO THE AUCTIONEERING CIRCUIT FOR THE 15VDC PWR SUPPLIES.Actions Taken Details: Previous Site ID: Site Change Flag: NOProblem Evaluation Report - Customer ImpactOutage
PER Vault Summary Report for PER: 3516 PER Number: 3516                                          Status: ARCHIVE                        Status Date:
Status By ID:
Site / Org: WBN                                  PER Level:                               CAP Due Date:
Unit:                              Date of Occurence:                         Long Lead CA Date:
Analysis Type:
PER Summary:
Problem Evaluation Report - User Information Originator ID:                                                                               On Behalf of ID:
Originator Name:                                                                              On Behalf of Name:
Originator Phone:                                                                            On Behalf of Phone:
Originator Email:                                                                            On Behalf of Email:
Problem Evaluation Report - Problem Details Initiating Department:                                                                                           Asset:
Site / Org: WBN                                                                                Asset Site:                         The source of the SBU:                                                                                       Location:
BU:                                                                   Additional Location Details:
scenario for the Unit:                                                                                  Plant System:                           question.
Corrected Immediately: NO                                                                                  Reported Date:
Date Of Occurence:
PER Summary:
PER Details: ONE 15VDC POWER SUPPLY FOR SSPS TR B,, LOCATED IN 1-R-50,, HAS GONE BAD. THIS BROUGHT IN A GENERAL WARNING FOR B TR SSPS. B TR SSPS IS STILL OPERABLE DUE TO THE AUCTIONEERING CIRCUIT FOR THE 15VDC PWR SUPPLIES.
Immediate Actions Taken:
Actions Taken Details:
Previous Site ID:
Site Change Flag: NO Problem Evaluation Report - Customer Impact Outage


==Reference:==
==Reference:==
Customer  
As Found Condition:              Customer  


==Reference:==
==Reference:==
As Found Condition: Problem Evaluation Report - Regulatory ImpactPotential Environmental Issue: NOPotential Safety Issue: NOPotential Operability Issue: Potential Reactivity Issue: Potential Reportability Issue: NONOPER Level: PER Category: Tier: Management Screening - Review Results / ApprovalGood Catch: Analysis Type: CAP Due Date: Responsible Org: Comments: Justification Details: Comment Details: NOPotential Margin Issue: NOJustification: Page 1 of 25Wednesday, December 10, 2014TVA RESTRICTED INFORMATIONThesourceofthescenarioforthe question.ONE 15VDC POWER SUPPLY FOR SSPS TR B, LOCATED IN 1-R-50, HAS GONE BAD. THIS BROUGHT IN A GENERAL WARNING FOR B TR SSPS. B,,TR SSPS IS STILL OPERABLE DUE TO THE AUCTIONEERING CIRCUIT FOR THE 15VDC PWR SUPPLIES.
 
PER Vault Summary Report for PER: 3516Event Number: Event
Problem Evaluation Report - Regulatory Impact Potential Environmental Issue: NO                Potential Operability Issue:       Potential Reactivity Issue: NO Potential Reportability Issue: NO                  Potential Safety Issue: NO Management Screening - Review Results / Approval PER Level:                        Tier:                                          PER Category:              Good Catch: NO Potential Margin Issue: NO Justification:
Justification Details:
Responsible Org:                                                                          Analysis Type:             CAP Due Date:
Comments:
Comment Details:
Page 1 of 25                                                                  TVA RESTRICTED INFORMATION                                          Wednesday, December 10, 2014
 
PER Vault Summary Report for PER: 3516 Regulatory Reviews - Environmental Review Event Number:         Event


== Description:==
== Description:==
Regulatory Reviews - Environmental ReviewEnvironmental Issue?: Environmental Type: Environmental Area: EMS Process: Source Code: Findings Code: Notice of Violation?: NOEvent Repeat?: NOJustification for Environ. Disposition: Reviewer ID: Review Date: Non-Conformance Code: Justification Details: Potential Environmental Issue?: NORegulatory Reviews - Safety ReviewSafety Event Number: Potential Safety?: NOSafety Issue?: Source Code: Findings Code: Violation Notice Required?: NOJustification for Safety Disposition: Reviewer ID: Review Date: Justification Details: Regulatory Reviews - Operations ReviewPotential Operability Issue?: Operability Issue?: Operability Actions: Operability Actions Details: Potential Reportability Issue?: Reportability Issue?: Operability Reviewer: Operability Review Date: Reportability Reviewer: Reportability Review Date: NONONONOEngineering Evaluation Needed?: Required Date: Ops Notified Other Sites?: NORegulatory Reviews - Engineering EvaluationSpecified (Safety) Function Maintained?: Evaluation Summary: CLB Affected: CLB Affected Details: Immediate/Compensatory Measures?: NONOEngineering Evaluation Needed?: Required Date: Evaluation Details: Outside CLB?: NOPage 3 of 25Wednesday, December 10, 2014TVA RESTRICTED INFORMATIONThecrewscreenedtheissue (correctly)asnot animpacttothe operabilityof SSPS.Potential Operability Issue?NO


Notanimpacttooperability.
Environmental Issue?:        Environmental Type:                        Event Repeat?: NO        Potential Environmental Issue?: NO Source Code:      Environmental Area:                  Notice of Violation?: NO Non-Conformance Code:              Findings Code:                        EMS Process:
NPG Standard Programs and Processes Temporary Modifications Temporary Configuration Changes NPG-SPP-09.5 Rev. 0009 Page 18 of 77 3.6 Procedurally Controlled Temporary Modifications (PCTM) (continued)    C. The procedure will include the following administrative information: 1. The section of the procedure that implements the PCTM will be clearly identified as a PCTM. 2. The section will include a note that any changes will require a 50.59 / 72.48 evaluation, a Technical Evaluation, and the Design Control review. (except for non-intent changes - minor or editorial)   3. If the PCTM is installed greater than one shift, the modification will be tagged and entered in the Temporary Modification Log. 3.7 Temporary Modifications in Support of Maintenance (WO-TM) A. Temporary modification in support of maintenance are implemented under 10 CFR 50.65 (Maintenance Rule) rather than 10 CFR 50.59 (Changes, Tests, and Experiments). The modification may remain installed for 90 days at power under 50.65. Beyond 90 days at power, a 50.59 / 72.48 review is required. B. The modification must meet the following criteria to be processed as a WO-TM: 1. The modification must be in direct support of maintenance (for example, necessary to establish work conditions or provide equipment necessary to perform work - see Attachment 19 for further clarification) 2. The modification must be controlled by an active WO.
Justification for Environ. Disposition:
: 3. The modification must not impact the decision-making capability of the plant Operators and must not require changes to operating procedures or Operator training. C. A Technical Evaluation (TE) review is required for all WO-TMs. 1. Engineering will perform a Technical Evaluation in accordance with the Technical Evaluation Form (Attachment 3) and the step-by-step instructions in Attachment
Justification Details:
: 21. 2. Engineering will provide the Technical Evaluation to Planning who will incorporate any special instructions or requirements into the WO package. 3. The Technical Evaluation will be included in the WO package. D. A 50.59 / 72.48 review is required under the following conditions: 1. If plant personnel expect the modification will be installed less than 90 days, then a 50.59 / 72.48 review is not required. 2. If plant personnel know beforehand that the modification will be in place for more than 90 days, then a 50.59 / 72.48 review is required. 3.7 Temporary Modifications in Support of Maintenance (WO-TM)C. A Technical Evaluation (TE) review is required for all WO-TMs. ThisisthecategoryofTmoddescribed bythequestion.Therefore,aTEis requiredforthe Tmodproposedby thequestion.
Reviewer ID:
: 87. Given the following conditions: - Unit 1 is in MODE 3. Subsequently: - A Loss of 120v AC Vital Instrument Power Board 1-III occurs. AOI-25.03, Loss of 120V AC Vital Instrument Power Boards 1-III or 2-III, directs use of 1-SOI-235.03, 120V AC Vital Power System 1-III, to restore the board Which ONE of the following describes the status of the system AND which procedure will allow exiting the T/S LCO? During the loss of the vital board, ____(1)____ Unit 1 SSPS Train A ESF relays COULD be energized. Implementing section ____(2)____ of 1-SOI-235.03 would re-energize the board and allows BOTH T/S LCOs 3.8.7, Inverters and 3.8.9, Distribution Systems - Operating to be exited. A. (1) ONLY the master (2) 8.1, Transferring 480V AC Vital Transfer Switch III to Alternate 480V Power Supply B. (1) BOTH the master and the slave (2) 8.1, Transferring 480V AC Vital Transfer Switch III to Alternate 480V Power Supply C. (1) ONLY the master (2) 8.3, Transfer 120V AC Vital Instrument Power Board 1-III to Spare 120V AC Vital Inverter 0-III D. (1) BOTH the master and the slave (2) 8.3, Transfer 120V AC Vital Instrument Power Board 1-III to Spare 120V AC Vital Inverter 0-III CORRECT ANSWER:D DISTRACTOR ANALYSIS:  A. Incorrect: As seen on the simplified Westinghouse drawing, the SSPS power supply distribution involves the provision of 120V AC Vital Instrumentation Buses I and III to the "A" train SSPS and Buses II and IV to the "B" train SSPS. One may observe on this simplified drawing that Buses I and III each power a 48V and 15V power supply pair. The 48V power supplies are auctioneered together to provide a 48V bus within the SSPS rack. One of the users of this bus's power is the operating coils for the SSPS master relays. Therefore, the loss of one of the two 120V AC Vital Instrument Buses which is provided to a train of SSPS will not cause a loss of power to the SSPS master relays because the remaining Vital Instrument Bus will continue to supply the redundant feed of 48V power. Another feed from Bus I to the "A" train SSPS is the provision of 120V AC power to the slave relay coils. One must note that there is no redundant supply of AC power to these slave relays. Therefore, if Bus I is lost, the "A" train SSPS slave relays will be depowered and thus unable to actuate. Germane to this question is that the loss of Bus III 120V AC will cause no impact to the slave relays. Therefore, upon the loss of Bus III 120V AC, both the master and slave relays could be energized.
Review Date:
Regulatory Reviews - Safety Review Safety Event Number:          Potential Safety?: NO      Violation Notice Required?: NO            Source Code:
Safety Issue?:          Findings Code:
Justification for Safety Disposition:
Justification Details:
Reviewer ID:
Review Date:
Regulatory Reviews - Operations Review Issue?
Potential Operability Issue?:  NO Operability Issue?: NO                                                                The crew screened Operability Actions:
the issue Operability Actions Details:
Potential Reportability Issue?:  NO (correctly) as not Reportability Issue?:   NO                                                                an impact to the Engineering Evaluation Needed?:                                                                        operability of Required Date:
SSPS.
Ops Notified Other Sites?:    NO Operability Reviewer:
Operability Review Date:
Reportability Reviewer:
Reportability Review Date:
Regulatory Reviews - Engineering Evaluation Engineering Evaluation Needed?:
Required Date:
Evaluation Summary:
Evaluation Details:
Outside CLB?: NO CLB Affected:
CLB Affected Details:
Specified (Safety) Function Maintained?: NO Immediate/Compensatory Measures?: NO Page 3 of 25                                                    TVA RESTRICTED INFORMATION                                      Wednesday, December 10, 2014


It is plausible to believe that the slave relays could not be energized provided that one believed that Bus III powered the slave relay coils.  
Not an impact to operability.


As seen in the print excerpt taken from 1-45W700-1, the 120V AC Vital Instrument buses are provided with power from an inverter unit. This unit can provide power from three basic sources: 1. power can be provided from an inverter which is supplied DC from the battery board, 2. power can be provided from an inverter which is supplied from an AC feed which is rectified and 3. power can be provided from a transformed and regulated AC feed which bypasses the inverter completely.  
NPG Standard                Temporary Modifications                NPG-SPP-09.5 Programs and          Temporary Configuration Changes              Rev. 0009 Processes                                                        Page 18 of 77 3.6      Procedurally Controlled Temporary Modifications (PCTM) (continued)
C. The procedure will include the following administrative information:
This is the category  1. The section of the procedure that implements the PCTM will be clearly identified of Tmod described          as a PCTM.
by the question.
: 2. The section will include a note that any changes will require a 50.59 / 72.48 evaluation, a Technical Evaluation, and the Design Control review. (except for non-intent changes - minor or editorial)
: 3. If the PCTM is installed greater than one shift, the modification will be tagged and entered in the Temporary Modification Log.
3.7      Temporary Modifications in Support of Maintenance (WO-TM)
A. Temporary modification in support of maintenance are implemented under 10 CFR 50.65 (Maintenance Rule) rather than 10 CFR 50.59 (Changes, Tests, and Experiments). The modification may remain installed for 90 days at power under 50.65. Beyond 90 days at power, a 50.59 / 72.48 review is required.
B. The modification must meet the following criteria to be processed as a WO-TM:
: 1. The modification must be in direct support of maintenance (for example, Therefore, a TE is necessary to establish work conditions or provide equipment necessary to required for the             perform work - see Attachment 19 for further clarification)
Tmod proposed by the question.          2. The modification must be controlled by an active WO.
: 3. The modification must not impact the decision-making capability of the plant Operators and must not require changes to operating procedures or Operator training.
C. A Technical Evaluation (TE) review is required for all WO-TMs.
: 1. Engineering will perform a Technical Evaluation in accordance with the Technical Evaluation Form (Attachment 3) and the step-by-step instructions in Attachment 21.
: 2. Engineering will provide the Technical Evaluation to Planning who will incorporate any special instructions or requirements into the WO package.
: 3. The Technical Evaluation will be included in the WO package.
D. A 50.59 / 72.48 review is required under the following conditions:
: 1. If plant personnel expect the modification will be installed less than 90 days, then a 50.59 / 72.48 review is not required.
: 2. If plant personnel know beforehand that the modification will be in place for more than 90 days, then a 50.59 / 72.48 review is required.


From the perspective of operability, one may see that T/S LCO 3.8.9 allows any one of the three aforementioned sources of power to supply a 120V AC bus. One may further see that T/S LCO 3.8.7 is different in that it does not allow the transformed an regulated AC feed to be utilized. Therefore, the use of section 8.1 (which would place the transformed an regulated AC feed in service) would NOT allow T/S LCO 3.8.7 to be satisfied. The plausibility to this distractor is lent by the fact that T/S LCO 3.8.9 allows the use of the bypass feed while T/S LCO 3.8.7 does not. 
87.
Given the following conditions:
    -  Unit 1 is in MODE 3.
Subsequently:
    -  A Loss of 120v AC Vital Instrument Power Board 1-III occurs.
    -  1-AOI-25.03, Loss of 120V AC Vital Instrument Power Boards 1-III or 2-III, directs use of 1-SOI-235.03, 120V AC Vital Power System 1-III, to restore the board Which ONE of the following describes the status of the system AND which procedure will allow exiting the T/S LCO?
During the loss of the vital board, ____(1)____ Unit 1 SSPS Train A ESF relays COULD be energized.
Implementing section ____(2)____ of 1-SOI-235.03 would re-energize the board and allows BOTH T/S LCOs 3.8.7, Inverters and 3.8.9, Distribution Systems - Operating to be exited.
A.    (1)    ONLY the master (2)    8.1, Transferring 480V AC Vital Transfer Switch III to Alternate 480V Power Supply B.    (1)    BOTH the master and the slave (2)   8.1, Transferring 480V AC Vital Transfer Switch III to Alternate 480V Power Supply C.   (1)    ONLY the master (2)    8.3, Transfer 120V AC Vital Instrument Power Board 1-III to Spare 120V AC Vital Inverter 0-III D.   (1)    BOTH the master and the slave (2)    8.3, Transfer 120V AC Vital Instrument Power Board 1-III to Spare 120V AC Vital Inverter 0-III


Note that both T/S LCO 3.8.7 and 3.8.9 are applicable in Modes 1-4. B. Incorrect: While it is correct that both the master and slave relays would be able to be energized, it is not correct and yet plausible that the use of the bypass AC feed would allow T/S LCO 3.8.7 to be MET. C. Incorrect: It is incorrect and yet plausible that only the master relays could be energized. It is correct (as discussed), that the transfer of the AC Vital Board to the spare inverter would allow both T/S LCO 3.8.9 and T/S LCO 3.8.7 to be exited. D. Correct: As noted, both of the parts of this answer are correct.
CORRECT ANSWER:                                                              D DISTRACTOR ANALYSIS:
Question Number: 87  Tier:  2 Group:  1 K/A: 013 Engineered Safety Features Actuation System A2 Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS; and (b) based Ability on those predictions, use procedures to Correct: control, or mitigate the consequences of those malfunctions or operations; Importance Rating: A2.04 Loss of instrument bus
A. Incorrect: As seen on the simplified Westinghouse drawing, the SSPS power supply distribution involves the provision of 120V AC Vital Instrumentation Buses I and III to the A train SSPS and Buses II and IV to the B train SSPS. One may observe on this simplified drawing that Buses I and III each power a 48V and 15V power supply pair.
The 48V power supplies are auctioneered together to provide a 48V bus within the SSPS rack. One of the users of this buss power is the operating coils for the SSPS master relays. Therefore, the loss of one of the two 120V AC Vital Instrument Buses which is provided to a train of SSPS will not cause a loss of power to the SSPS master relays because the remaining Vital Instrument Bus will continue to supply the redundant feed of 48V power. Another feed from Bus I to the A train SSPS is the provision of 120V AC power to the slave relay coils. One must note that there is no redundant supply of AC power to these slave relays. Therefore, if Bus I is lost, the A train SSPS slave relays will be depowered and thus unable to actuate. Germane to this question is that the loss of Bus III 120V AC will cause no impact to the slave relays. Therefore, upon the loss of Bus III 120V AC, both the master and slave relays could be energized.
It is plausible to believe that the slave relays could not be energized provided that one believed that Bus III powered the slave relay coils.
As seen in the print excerpt taken from 1-45W700-1, the 120V AC Vital Instrument buses are provided with power from an inverter unit. This unit can provide power from three basic sources: 1. power can be provided from an inverter which is supplied DC from the battery board,
: 2. power can be provided from an inverter which is supplied from an AC feed which is rectified and 3. power can be provided from a transformed and regulated AC feed which bypasses the inverter completely.
From the perspective of operability, one may see that T/S LCO 3.8.9 allows any one of the three aforementioned sources of power to supply a 120V AC bus. One may further see that T/S LCO 3.8.7 is different in that it does not allow the transformed an regulated AC feed to be utilized. Therefore, the use of section 8.1 (which would place the transformed an regulated AC feed in service) would NOT allow T/S LCO 3.8.7 to be satisfied. The plausibility to this distractor is lent by the fact that T/S LCO 3.8.9 allows the use of the bypass feed while T/S LCO 3.8.7 does not.
Note that both T/S LCO 3.8.7 and 3.8.9 are applicable in Modes 1-4.
B. Incorrect: While it is correct that both the master and slave relays would be able


10 CFR Part 55: (CFR: 41.5 / 43.5 / 45.3 / 45.13)  10CFR55.43.b: 10 CFR 55.43(b)(2)
to be energized, it is not correct and yet plausible that the use of the bypass AC feed would allow T/S LCO 3.8.7 to be MET.
C. Incorrect: It is incorrect and yet plausible that only the master relays could be energized. It is correct (as discussed), that the transfer of the AC Vital Board to the spare inverter would allow both T/S LCO 3.8.9 and T/S LCO 3.8.7 to be exited.
D. Correct: As noted, both of the parts of this answer are correct.


K/A Match: The K/A is matched because the question requires predicting the impact of the loss if an instrument bus on the ESFAS. The question then requires selecting the correct section of an SOI to both restore power and maintain T/S LCO operability. Technical  
Question Number:        87 Tier:    2  Group:        1 K/A:    013 Engineered Safety Features Actuation System A2 Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS; and (b) based Ability on those predictions, use procedures to Correct: control, or mitigate the consequences of those malfunctions or operations; Importance Rating:      A2.04 Loss of instrument bus 10 CFR Part 55:        (CFR: 41.5 / 43.5 / 45.3 / 45.13) 10CFR55.43.b:        10 CFR 55.43(b)(2)
K/A Match:   The K/A is matched because the question requires predicting the impact of the loss if an instrument bus on the ESFAS. The question then requires selecting the correct section of an SOI to both restore power and maintain T/S LCO operability.
Technical  


==Reference:==
==Reference:==
1-45W700-1 Simplified Westinghouse graphic showing the Power Distribution to SSPS 1-SOI-235.03, 120V AC Vital Power System 1 III T/S Basis for LCO 3.8.7, Inverters - Operating T/S Basis for LCO 3.8.9, Distribution Systems - Operating Proposed references to be provided: None  Learning Objective: 3-OT-SYS235A 4. EXPLAIN the physical connections and/or cause-effect relationships between the 120 Volt AC System and the following: a. Solid State Protection System (SSPS)  
1-45W700-1 Simplified Westinghouse graphic showing the Power Distribution to SSPS 1-SOI-235.03, 120V AC Vital Power System 1 III T/S Basis for LCO 3.8.7, Inverters - Operating T/S Basis for LCO 3.8.9, Distribution Systems - Operating Proposed references to     None be provided:
: 10. DESCRIBE the following aspects of TS and TRs
Learning Objective:         3-OT-SYS235A
: b. The Limiting Conditions for Operation, Applicability, and Bases. Cognitive Level:     Higher X Lower     Question Source:     New   Modified Bank   Bank   X   Question History: Bank Question 013 A2.04 87 which was used on the 06/2011 WBN NRC exam. Comments: The question meets the general SRO only criteria of "Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations."   
: 4. EXPLAIN the physical connections and/or cause-effect relationships between the 120 Volt AC System and the following:
: a. Solid State Protection System (SSPS)
: 10. DESCRIBE the following aspects of TS and TRs
: b. The Limiting Conditions for Operation, Applicability, and Bases.
Cognitive Level:
Higher               X Lower Question Source:
New Modified Bank Bank                   X Question History:           Bank Question 013 A2.04 87 which was used on the 06/2011 WBN NRC exam.
Comments:                   The question meets the general SRO only criteria of Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.
 
SSPS Power Supply Distribution Bus I                                120V AC Vital Instrumentation Bus II Bus III Bus IV From Train A                          From Train B Safeguards                              Safeguards Train A            Test Panel            Train B          Test Panel I                                          I 48v 15v    To Slave                                To Slave Relays                                  Relays II                                          II 48v 15v III                                        III 48v 15v IV                                          IV 48v 15v


Bus IBus IIBus IIIBusIVTo SlaveRelaysTo SlaveRelaysSSPS Power Supply DistributionTrain ATrain B120V AC Vital Instrumentation48v15v48v15vIIIIIIIVIIIIIIIV48v15v15v48vFrom Train 'A' Safeguards Test PanelFrom Train 'B' Safeguards Test Panel Distribution Systems - Operating B 3.8.9 BASES    (continued)    Watts Bar-Unit 1 B 3.8-91 Revision 67, 75, 76, 77, 78        LCO   Maintaining the Train A and Train B AC, four channels of vital DC, and   (continued) four channels of AC vital bus electrical power distribution subsystems OPERABLE ensures that the redundancy incorporated into the design of ESF is not defeated. Therefore, a single failure within any system or within the electrical power distribution subsystems will not prevent safe shutdown of the reactor. OPERABLE AC electrical power distribution subsystems require the associated buses, load centers, motor control centers, and distribution panels to be energized to their proper voltages. OPERABLE DC electrical power distribution subsystems require the associated buses to be energized to their proper voltage from either the associated battery or charger. OPERABLE vital bus electrical power distribution subsystems require the associated buses to be energized to their proper voltage from the associated unit or spare inverter via inverted DC voltage, unit or spare inverter using internal AC source, or the regulated transformer bypass source. In addition, tie breakers between redundant safety related AC, vital DC, and AC vital bus power distribution subsystems, if they exist, must be open. This prevents any electrical malfunction in any power distribution subsystem from propagating to the redundant subsystem, that could cause the failure of a redundant subsystem and a loss of essential safety function(s). If any tie breakers are closed, the affected redundant electrical power distribution subsystems are considered inoperable. This applies to the onsite, safety related redundant electrical power distribution subsystems. It does not, however, preclude redundant 6.9 kV shutdown boards from being powered from the same offsite circuit. APPLICABILITY The electrical power distribution subsystems are required to be OPERABLE in MODES 1, 2, 3, and 4 to ensure that:     a. Acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result of AOOs or abnormal transients; and     b. Adequate core cooling is provided, and containment OPERABILITY and other vital functions are maintained in the event of a postulated DBA. Electrical power distribution subsystem requirements for MODES 5 and 6 are covered in the Bases for LCO 3.8.10, "Distribution Systems - Shutdown."
From an inverter                                                                  Distribution Systems - Operating (normal or spare)                                                                                            B 3.8.9 from either 1. the inverter whenBASES powered by the LCO             Maintaining the Train A and Train B AC, four channels of vital DC, and battery, 2. the (continued)     four channels of AC vital bus electrical power distribution subsystems inverter when                OPERABLE ensures that the redundancy incorporated into the design of ESF is powered by the ac to          not defeated. Therefore, a single failure within any system or within the electrical dc converter, or 3.          power distribution subsystems will not prevent safe shutdown of the reactor.
Inverters - Operating B 3.8.7  BASES    (continued)   Watts Bar-Unit 1 B 3.8-82 Revision 58, 67, 75, 76, 77, 78,    97  Amendment 45, 76  APPLICABLE  Inverters are a part of the distribution systems and, as such, satisfy Criterion 3 SAFETY ANALYSIS of the NRC Policy Statement.
the isolimiter OPERABLE AC electrical power distribution subsystems require the associated (regulated                    buses, load centers, motor control centers, and distribution panels to be transformer bypass            energized to their proper voltages. OPERABLE DC electrical power distribution source)                      subsystems require the associated buses to be energized to their proper voltage from either the associated battery or charger. OPERABLE vital bus electrical power distribution subsystems require the associated buses to be energized to their proper voltage from the associated unit or spare inverter via inverted DC THIS IS DIFFERENT          voltage, unit or spare inverter using internal AC source, or the regulated THAN T/S LCO 3.8.7        transformer bypass source.
  (continued)
In addition, tie breakers between redundant safety related AC, vital DC, and AC vital bus power distribution subsystems, if they exist, must be open. This prevents any electrical malfunction in any power distribution subsystem from propagating to the redundant subsystem, that could cause the failure of a redundant subsystem and a loss of essential safety function(s). If any tie breakers are closed, the affected redundant electrical power distribution subsystems are considered inoperable. This applies to the onsite, safety related redundant electrical power distribution subsystems. It does not, however, preclude redundant 6.9 kV shutdown boards from being powered from the same offsite circuit.
APPLICABILITY   The electrical power distribution subsystems are required to be OPERABLE in MODES 1, 2, 3, and 4 to ensure that:
: a.       Acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result of AOOs or abnormal transients; and
: b.       Adequate core cooling is provided, and containment OPERABILITY and other vital functions are maintained in the event of a postulated DBA.
Electrical power distribution subsystem requirements for MODES 5 and 6 are covered in the Bases for LCO 3.8.10, "Distribution Systems - Shutdown."
(continued)
Watts Bar-Unit 1                             B 3.8-91                      Revision 67, 75, 76, 77, 78


LCO   The inverters ensure the availability of AC electrical power for the systems instrumentation required to shut down the reactor and maintain it in a safe condition after an anticipated operational occurrence (A00) or a postulated DBA.
Inverters - Operating B 3.8.7 BASES APPLICABLE      Inverters are a part of the distribution systems and, as such, satisfy Criterion 3 SAFETY ANALYSIS  of the NRC Policy Statement.
Maintaining the required inverters OPERABLE ensures that the redundancy incorporated into the design of the RPS and ESFAS instrumentation and controls is maintained. The twelve inverters (one Unit 1, one Unit 2 and one spare per channel) ensure an uninterruptible supply of AC electrical power to the AC vital buses even if the 6.9 kV shutdown boards are de-energized. OPERABLE inverters require the associated AC vital bus to be powered by an inverter with output voltage and frequency within tolerances and power input to the inverter from a 125 VDC vital battery. Alternatively, power supply may be from an internal AC source via rectifier as long as the vital battery is available as the uninterruptible power supply. The unit inverters have an associated bypass supply provided by a regulated transformer that is automatically connected to the associated AC vital bus in the event of inverter failure or overload. The bypass supply is not battery-backed and thus does not meet requirements for inverter operability. The spare inverters do not have an associated bypass supply. Additionally, the inverter channel must not be connected to the cross train 480 V power supply.
(continued)
APPLICABILITY The inverters are required to be OPERABLE in MODES 1, 2, 3, and 4 to ensure that:
LCO             The inverters ensure the availability of AC electrical power for the systems instrumentation required to shut down the reactor and maintain it in a safe condition after an anticipated operational occurrence (A00) or a postulated DBA.
: a. Acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result of AOOs or abnormal transients; and
Maintaining the required inverters OPERABLE ensures that the redundancy The inverter is              incorporated into the design of the RPS and ESFAS instrumentation and controls operable when the            is maintained. The twelve inverters (one Unit 1, one Unit 2 and one spare per inverter is                  channel) ensure an uninterruptible supply of AC electrical power to the AC vital buses even if the 6.9 kV shutdown boards are de-energized.
: b. Adequate core cooling is provided, and containment OPERABILITY and other vital functions are maintained in the event of a postulated DBA.
supplying the vital AC board (not the            OPERABLE inverters require the associated AC vital bus to be powered by an regulated bypass            inverter with output voltage and frequency within tolerances and power input to supply).                    the inverter from a 125 VDC vital battery. Alternatively, power supply may be from an internal AC source via rectifier as long as the vital battery is available as the uninterruptible power supply. The unit inverters have an associated bypass supply provided by a regulated transformer that is automatically connected to the associated AC vital bus in the event of inverter failure or overload. The bypass supply is not battery-backed and thus does not meet requirements for inverter operability. The spare inverters do not have an associated bypass supply.
Additionally, the inverter channel must not be connected to the cross train 480 V power supply.
APPLICABILITY   The inverters are required to be OPERABLE in MODES 1, 2, 3, and 4 to ensure that:
: a.       Acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result of AOOs or abnormal transients; and
: b.       Adequate core cooling is provided, and containment OPERABILITY and other vital functions are maintained in the event of a postulated DBA.
Inverter requirements for MODES 5 and 6 are covered in the Bases for LCO 3.8.8, "Inverters - Shutdown."
Inverter requirements for MODES 5 and 6 are covered in the Bases for LCO 3.8.8, "Inverters - Shutdown."
WBN Unit 1 120V AC Vital Power System 1-III 1-SOI-235.03 Rev. 0000 Page 25 of 35   Date________     Initials     8.0 INFREQUENT OPERATIONS 8.1 Transferring 480V AC Vital Transfer Switch III to Alternate 480V Power Supply CAUTIONS 1) Consideration should be given to a possible loss of Channel 3 SSPS and ESF should 120V AC Vital Power Board 1-III and 2-III lose potential. 2) EMERGENCY feeder from 480V SHUTDOWN BOARD 2B1-B on 0-XSW-236-3, 480V AC VITAL TRANSFER SWITCH III, is NOT accounted for in D/G loading calculations and shall NOT be used without engineering evaluation (see note 9, drawing 1-15E500-2). [1] OBTAIN current approved engineering evaluation for this performance and attach a copy to this Data Package. ________ SRO [2] CHECK 1-EI-235-3/V2, BATTERY INPUT on Inverter 1-III OR 0-EI-235-3/V2, BATTERY INPUT on Inverter 0-III, if in service, to be 133-140 VDC. ________ [3] IF INSTRUMENT POWER A RACK TRANSFER SWITCH, in NORMAL FEEDER, THEN PLACE INSTRUMENT POWER A RACK TRANSFER SWITCH, in ALTERNATE FEEDER. [2-M-7] ________  ________
(continued)
CV [4] IF INSTRUMENT POWER B RACK TRANSFER SWITCH, in ALTERNATE FEEDER, THEN PLACE INSTRUMENT POWER B RACK TRANSFER SWITCH, in NORMAL FEEDER. [2-M-7] ________
Watts Bar-Unit 1                           B 3.8-82                  Revision 58, 67, 75, 76, 77, 78, 97 Amendment 45, 76
________
 
CV WBN Unit 1 120V AC Vital Power System 1-III 1-SOI-235.03 Rev. 0000 Page 26 of 35   Date________     Initials   8.1 Transferring 480V AC Vital Transfer Switch III to Alternate 480V Power Supply (continued)    [5] EVALUATE possible effects on all feeds from 120V Instrument Power Distribution Panel 2-A, including feeds from Panel 2-M-7, due to momentary loss of potential, to include the following systems:  
WBN              120V AC Vital Power System 1-III       1-SOI-235.03 Unit 1                                                  Rev. 0000 Page 25 of 35 Date________                                                                 Initials This section renders T/S LCO 3.8.7 NOT MET 8.0   INFREQUENT OPERATIONS 8.1   Transferring 480V AC Vital Transfer Switch III to Alternate 480V Power Supply CAUTIONS
: 1) Consideration should be given to a possible loss of Channel 3 SSPS and ESF should 120V AC Vital Power Board 1-III and 2-III lose potential.
: 2) EMERGENCY feeder from 480V SHUTDOWN BOARD 2B1-B on 0-XSW-236-3, 480V AC VITAL TRANSFER SWITCH III, is NOT accounted for in D/G loading calculations and shall NOT be used without engineering evaluation (see note 9, drawing 1-15E500-2).
[1]   OBTAIN current approved engineering evaluation for this performance and attach a copy to this Data Package.               ________
SRO
[2]   CHECK 1-EI-235-3/V2, BATTERY INPUT on Inverter 1-III OR 0-EI-235-3/V2, BATTERY INPUT on Inverter 0-III, if in service, to be 133-140 VDC.                                               ________
[3]   IF INSTRUMENT POWER A RACK TRANSFER SWITCH, in NORMAL FEEDER, THEN PLACE INSTRUMENT POWER A RACK TRANSFER SWITCH, in ALTERNATE FEEDER. [2-M-7]                             ________
CV
[4]   IF INSTRUMENT POWER B RACK TRANSFER SWITCH, in ALTERNATE FEEDER, THEN PLACE INSTRUMENT POWER B RACK TRANSFER SWITCH, in NORMAL FEEDER. [2-M-7]                                 ________
CV
 
WBN            120V AC Vital Power System 1-III        1-SOI-235.03 Unit 1                                                  Rev. 0000 Page 26 of 35 Date________                                                            Initials 8.1  Transferring 480V AC Vital Transfer Switch III to Alternate 480V Power Supply (continued)
[5]  EVALUATE possible effects on all feeds from 120V Instrument Power Distribution Panel 2-A, including feeds from Panel 2-M-7, due to momentary loss of potential, to include the following systems:
Access control system -Nuclear Security- Momentary loss if on alternate supply (notify Nuclear Security).                    ________
[6]  CHECK status windows for other ESF or SSPS channels LIT which could cause a Reactor Trip or Safety Injection should channel 1 be lost on the power supply transfer.                ________
[7]  CLOSE 0-BKR-236-3A, ALT FDR FOR VITAL BATT CHGR III (0-CHGR-236-3), on 480V SHUTDOWN BOARD 2B1-B
[C/10A].                                                        ________
CV
[8]  CHECK 0-XSW-236-3, 480V AC VITAL TRANSFER SWITCH III, EMERGENCY (supply feeder) red light ON.            ________
[9]  PLACE 0-XSW-236-3, 480V AC VITAL TRANSFER SWITCH III, NORMAL supply to OFF.                              ________
CV
[10]  PLACE 0-XSW-236-3, 480V AC VITAL TRANSFER SWITCH III, EMERGENCY supply to ON.                            ________
CV
 
WBN           120V AC Vital Power System 1-III       1-SOI-235.03 Unit 1                                                Rev. 0000 Page 27 of 35 Date________                                                         Initials 8.1   Transferring 480V AC Vital Transfer Switch III to Alternate 480V Power Supply (continued)
[11]  CHECK the following equipment ENERGIZED:
[11.1]  120V AC VITAL INVERTER 1-III or 0-III, if in service  ________
[11.2]  120V AC VITAL INSTR POWER BOARD 1-III.                ________
[11.3]  120V AC VITAL INSTR POWER BOARD 2-III.                ________
[11.4]  125-V VITAL BATTERY CHARGER III.                      ________
[12]  PLACE INSTRUMENT POWER A RACK TRANSFER SWITCH, in NORMAL FEEDER. [2-M-7]                          ________
CV
[13]  CHECK 0-XSW-236-3, 480 V AC VITAL TRANSFER SWITCH III, NORMAL supply to OFF.                          ________
IV End of Section
 
WBN              120V AC Vital Power System 1-III          1-SOI-235.03 Unit 1                                                      Rev. 0000 Page 28 of 35 Date________                                                              Initials 8.2     Transferring 480V AC Vital Transfer Switch III to Normal 480V Power Supply CAUTION Consideration should be given to a possible loss of Channel 3 SSPS and ESF should 120V AC Vital Power Board 1-III and 2-III lose potential.
[1]    OBTAIN SRO approval prior to performing this Section.          ________
SRO
[2]    CHECK 1-EI-235-3/V2, BATTERY INPUT on Inverter 1-III OR 0-EI-235-3/V2, BATTERY INPUT on Inverter 0-III, if in service, to be 133-140 VDC.                                              ________
[3]    IF INSTRUMENT POWER A RACK TRANSFER SWITCH, in NORMAL FEEDER, THEN PLACE INSTRUMENT POWER A RACK TRANSFER SWITCH, in ALTERNATE FEEDER. [2-M-7]                            ________
CV
[4]    IF INSTRUMENT POWER B RACK TRANSFER SWITCH, in ALTERNATE FEEDER, THEN PLACE INSTRUMENT POWER B RACK TRANSFER SWITCH, in NORMAL FEEDER. [2-M-7]                              ________
CV
[5]   EVALUATE possible effects on all feeds from 120V Instrument Power Distribution Panel 2-A, including feeds from Panel 2-M-7, due to temporary loss of potential, to include the following systems:
Access control system -Nuclear Security- Momentary loss if on alternate supply (notify Nuclear Security).                    ________
[6]    CHECK status windows for other ESF or SSPS channels LIT which could cause a Reactor Trip or Safety Injection should channel III be lost on the power supply transfer.              ________
 
WBN            120V AC Vital Power System 1-III      1-SOI-235.03 Unit 1                                                Rev. 0000 Page 29 of 35 Date________                                                        Initials 8.2  Transferring 480V AC Vital Transfer Switch III to Normal 480V Power Supply (continued)
[7]    CHECK 0-XSW-236-3, 480V AC VITAL TRANSFER SWITCH III, NORMAL (supply feeder) red light ON.          ________
[8]    PLACE 0-XSW-236-3, 480 V AC VITAL TRANSFER SWITCH III, EMERGENCY supply to OFF.                      ________
CV
[9]    PLACE 0-XSW-236-3, 480 V AC VITAL TRANSFER SWITCH III, NORMAL supply to ON.                          ________
CV
[10]  CHECK the following equipment ENERGIZED:
[10.1]  120V AC VITAL INVERTER 1-III or 0-III, if in service. ________
[10.2]  120V AC VITAL INSTR POWER BOARD 1-III.                ________
[10.3]  120V AC VITAL INSTR POWER BOARD 2-III.                ________
[10.4]  125-V VITAL BATTERY CHARGER III.                      ________
[11]  PLACE INSTRUMENT POWER A RACK TRANSFER SWITCH, in NORMAL FEEDER. [2-M-7]                          ________
CV
[12]  OPEN 0-BKR-236-3A, ALT FDR FOR VITAL BATT CHGR III (0-CHGR-236-3), on 480V SHUTDOWN BOARD 2B1-B
[C/10A].                                                  ________
CV
[13]  CHECK 0-XSW-236-3, 480 V AC VITAL TRANSFER SWITCH III, EMERGENCY supply to OFF.                      ________
IV End of Section


Access control system -Nuclear Security- Momentary loss if on alternate supply (notify Nuclear Security). ________ [6] CHECK status windows for other ESF or SSPS channels LIT which could cause a Reactor Trip or Safety Injection should channel 1 be lost on the power supply transfer. ________ [7] CLOSE 0-BKR-236-3A, ALT FDR FOR VITAL BATT CHGR III (0-CHGR-236-3), on 480V SHUTDOWN BOARD 2B1-B [C/10A]. ________
WBN               120V AC Vital Power System 1-III       1-SOI-235.03 Unit 1                                                      Rev. 0000 Page 30 of 35 Date________                                                           This section allows both Initials T/S LCO 3.8.7 and T/S 8.3      Transfer 120V AC Vital Instrument Power Board 1-III to Spare          LCO 3.8.9 to be MET.
________
120V AC Vital Inverter 0-III NOTE This procedure section transfers the 120V AC power supply to 120 V AC Vital Instrument Power Board 1-III from 120V AC Vital Inverter 1-III to Spare 120V AC Vital Inverter 0-III.
CV [8] CHECK 0-XSW-236-3, 480V AC VITAL TRANSFER SWITCH III, EMERGENCY (supply feeder) red light ON. ________ [9] PLACE 0-XSW-236-3, 480V AC VITAL TRANSFER SWITCH III, NORMAL supply to OFF. ________  ________  CV [10] PLACE 0-XSW-236-3, 480V AC VITAL TRANSFER SWITCH III, EMERGENCY supply to ON. ________
This section is not used to energize a dead board. Refer to Sections 5.2 and 5.3 to energize dead 120V AC Vital Instrument Power Board 1-III using Spare 120 V AC Vital Inverter 0-III.
________
[1]     ENSURE Spare Inverter 0-III has been placed in service per Section 5.2, Startup of 120V AC Vital Inverter 0-III.              ________
CV WBN Unit 1 120V AC Vital Power System 1-III 1-SOI-235.03 Rev. 0000 Page 27 of 35   Date________     Initials   8.1 Transferring 480V AC Vital Transfer Switch III to Alternate 480V Power Supply (continued)    [11] CHECK the following equipment ENERGIZED: [11.1] 120V AC VITAL INVERTER 1-III or 0-III, if in service ________ [11.2] 120V AC VITAL INSTR POWER BOARD 1-III. ________
CV
[11.3] 120V AC VITAL INSTR POWER BOARD 2-III. ________
[2]     ENSURE 120V AC VITAL INVERTER SUPPLY AVAILABLE amber light on 120V AC VITAL INSTR POWER BD 1-III LIT.             ________
[11.4] 125-V VITAL BATTERY CHARGER III. ________ [12] PLACE INSTRUMENT POWER A RACK TRANSFER SWITCH, in NORMAL FEEDER. [2-M-7] ________
CV
________
[3]    ENSURE 120V AC ALTERNATE SUPPLY AVAILABLE amber light on 120V AC VITAL INSTR POWER BD 1-III LIT.                   ________
CV [13] CHECK 0-XSW-236-3, 480 V AC VITAL TRANSFER SWITCH III, NORMAL supply to OFF. ________
CV
IV  End of Section WBN Unit 1 120V AC Vital Power System 1-III 1-SOI-235.03 Rev. 0000 Page 28 of 35  Date________     Initials    8.2 Transferring 480V AC Vital Transfer Switch III to Normal 480V Power Supply  CAUTION Consideration should be given to a possible loss of Channel 3 SSPS and ESF should 120V AC Vital Power Board 1-III and 2-III lose potential. [1] OBTAIN SRO approval prior to performing this Section. ________ SRO [2] CHECK 1-EI-235-3/V2, BATTERY INPUT on Inverter 1-III OR 0-EI-235-3/V2, BATTERY INPUT on Inverter 0-III, if in service, to be 133-140 VDC. ________ [3] IF INSTRUMENT POWER A RACK TRANSFER SWITCH, in NORMAL FEEDER, THEN PLACE INSTRUMENT POWER A RACK TRANSFER SWITCH, in ALTERNATE FEEDER. [2-M-7] ________  ________
[4]     ENSURE 120V AC VITAL INVERTER & ALT SUPPLY IN SYNC blue light on 120V AC VITAL INSTR POWER BD 1-III LIT.                                                               ________
CV [4] IF INSTRUMENT POWER B RACK TRANSFER SWITCH, in ALTERNATE FEEDER, THEN  PLACE INSTRUMENT POWER B RACK TRANSFER SWITCH, in NORMAL FEEDER. [2-M-7] ________
CV
________
[5]     PLACE 1-XSW-235-3, 120V AC VITAL INSTR POWER BD 1-III TRANSFER, on 120V AC VITAL INSTR POWER BD 1-III to ALTERNATE.                                             ________
CV [5] EVALUATE possible effects on all feeds from 120V Instrument Power Distribution Panel 2-A, including feeds from Panel 2-M-7, due to temporary loss of potential, to include the following systems:
CV
[6]     CHECK 120V AC VITAL INSTR POWER BD 1-III ENERGIZED.                                                         ________


Access control system -Nuclear Security- Momentary loss if on alternate supply (notify Nuclear Security). ________ [6] CHECK status windows for other ESF or SSPS channels LIT which could cause a Reactor Trip or Safety Injection should channel III be lost on the power supply transfer. ________
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
WBN Unit 1 120V AC Vital Power System 1-III 1-SOI-235.03 Rev. 0000 Page 29 of 35  Date________    Initials    8.2 Transferring 480V AC Vital Transfer Switch III to Normal 480V Power Supply (continued)    [7] CHECK 0-XSW-236-3, 480V AC VITAL TRANSFER SWITCH III, NORMAL (supply feeder) red light ON. ________ [8] PLACE 0-XSW-236-3, 480 V AC VITAL TRANSFER SWITCH III, EMERGENCY supply to OFF. ________  ________
II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]:
CV [9] PLACE 0-XSW-236-3, 480 V AC VITAL TRANSFER SWITCH III, NORMAL supply to ON. ________
A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)]
________
Some examples of SRO exam items for this topic include:
CV [10] CHECK the following equipment ENERGIZED: [10.1] 120V AC VITAL INVERTER 1-III or 0-III, if in service. ________ [10.2] 120V AC VITAL INSTR POWER BOARD 1-III. ________
[10.3] 120V AC VITAL INSTR POWER BOARD 2-III. ________
[10.4] 125-V VITAL BATTERY CHARGER III. ________ [11] PLACE INSTRUMENT POWER A RACK TRANSFER SWITCH, in NORMAL FEEDER.  [2-M-7] ________
________
CV [12] OPEN 0-BKR-236-3A, ALT FDR FOR VITAL BATT CHGR III (0-CHGR-236-3), on 480V SHUTDOWN BOARD 2B1-B [C/10A]. ________
________
CV [13] CHECK 0-XSW-236-3, 480 V AC VITAL TRANSFER SWITCH III, EMERGENCY supply to OFF. ________  IV  End of Section WBN Unit 1 120V AC Vital Power System 1-III 1-SOI-235.03 Rev. 0000 Page 30 of 35  Date________    Initials    8.3 Transfer 120V AC Vital Instrument Power Board 1-III to Spare 120V AC Vital Inverter 0-III  NOTE This procedure section transfers the 120V AC power supply to 120 V AC Vital Instrument Power Board 1-III from 120V AC Vital Inverter 1-III to Spare 120V AC Vital Inverter 0-III. This section is not used to energize a dead board. Refer to Sections 5.2 and 5.3 to energize dead 120V AC Vital Instrument Power Board 1-III using Spare 120 V AC Vital Inverter 0-III. [1] ENSURE Spare Inverter 0-III has been placed in service per Section 5.2, Startup of 120V AC Vital Inverter 0-III. ________  ________
CV [2] ENSURE 120V AC VITAL INVERTER SUPPLY AVAILABLE amber light on 120V AC VITAL INSTR POWER BD 1-III LIT. ________
________
CV [3] ENSURE 120V AC ALTERNATE SUPPLY AVAILABLE amber light on 120V AC VITAL INSTR POWER BD 1-III LIT. ________  ________
CV [4] ENSURE 120V AC VITAL INVERTER & ALT SUPPLY IN SYNC blue light on 120V AC VITAL INSTR POWER BD 1-III LIT. ________
________
CV [5] PLACE 1-XSW-235-3, 120V AC VITAL INSTR POWER BD 1-III TRANSFER, on 120V AC VITAL INSTR POWER BD 1-III to ALTERNATE. ________
________
CV [6] CHECK 120V AC VITAL INSTR POWER BD 1-III ENERGIZED. ________
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 3 of 16  II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]: A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)] Some examples of SRO exam items for this topic include:
* Reporting requirements when the maximum licensed thermal power output is exceeded.
* Reporting requirements when the maximum licensed thermal power output is exceeded.
* Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems, fire doors, etc.
* Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems, fire doors, etc.
* The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
* The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
* National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
* National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
* Processes for TS and FSAR changes. Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic. B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)] Some examples of SRO exam items for this topic include:
* Processes for TS and FSAR changes.
* Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic.
B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]
Some examples of SRO exam items for this topic include:
The question requires the
* Application of Required Actions (Section 3) and Surveillance knowledge of the        Requirements (SR) (Section 4) in accordance with rules of application basis for T/S LCO        requirements (Section 1).
3.8.7 and 3.8.9.
* Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
* Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
* Knowledge of TS bases that are required to analyze TS required actions and terminology.
* Knowledge of TS bases that are required to analyze TS required actions and terminology.
* Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).
* Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of  1 hour action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.  
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of  1 hour action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. ROs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Page 3 of 16


SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. RO's are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 4 of 16 ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)         Above this line RO knowledge Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 5 of 16 Figure 1:  Screening for SRO-only linked to 10 CFR 55.43(b)(2) (Tech Specs)       
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)
RO knowledge Above this line Page 4 of 16


Can question be answered solely by knowing  1 hour TS/TRM Action? RO question YesNoCan question be answered solely by knowing the LCO/TRM information listed "above-the-line?" YesRO question NoCan question be answered solely by knowing the TS Safety Limits? YesRO question NoDoes the question involve one or more of the following for TS, TRM, or ODCM?
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)
(Tech Specs)
Can question be answered solely by knowing  1       Yes hour TS/TRM Action?                                           RO question No Can question be answered solely by knowing the       Yes LCO/TRM information listed above-the-line?                   RO question No Can question be answered solely by knowing the       Yes TS Safety Limits?                                             RO question No Does the question involve one or more of the following for TS, TRM, or ODCM?
* Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
* Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)
* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)                             Yes      SRO-only
* Knowledge of TS bases that is required to analyze TS required actions and terminology SRO-only question YesNoQuestion might not be linked to 10 CFR 55.43(b)(2) for SRO-only
* Knowledge of TS bases that is required to analyze TS                   question required actions and terminology No Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only Page 5 of 16
: 88. Given the following conditions: - Unit 1 is at 100% power. - The following equipment is INOPERABLE: 1. The TDAFWP room's DC emergency exhaust fan.  
 
: 2. The TDAFWP room's AC emergency exhaust fan.  
88.
: 3. 1-LCV-3-164, MDAFWP SG 1 SUPPLY. Of the THREE equipment items listed above, which ONE of the following lists the items that are required to be OPERABLE in accordance with T/S LCO 3.7.5, AFW System? A. ONLY 3 B. ONLY 2 and 3 C. ONLY 1 and 2 D. ONLY 1 and 3 CORRECT ANSWER:D DISTRACTOR ANALYSIS:   A. Incorrect: As seen in WBN-SDD-N3-30AB-4001, "Auxiliary Building Heating, Ventilation, Air Conditioning System," the TDAFW pump rooms are normally ventilated by the AB air exhaust system. Two 100% emergency exhaust fans, one AC operated and one DC operated, are provided in each TDAFW pump room-The DC-operated fan is installed to provide the required cooling in the event of a loss of all AC power, and will automatically start upon the start of the TDAFW pump. This is the only AFW pump available during a loss of all onsite and offsite AC power. The DC fan is the only means available to maintain the temperature requirements in the room. The DC fan is therefore a primary safety-related system component-The AC fan does not serve a safety-related function."  Therefore it is correct that in accordance with the AB HVAC system description, the TDAFWP room's DC emergency exhaust fan is safety-related.
Given the following conditions:
The basis for T/S LCO 3.7.5 reflects that: "This requires that the two motor driven AFW pumps be OPERABLE in two diverse paths, each supplying AFW to separate steam generators."  Therefore, each MDAFWP must be able to provide two S/Gs with AFW (or stated, the MDAFWPs must be able to provide all of the S/Gs). If one of the LCVs is INOPERABLE, then that train of AFW is rendered INOPERABLE. B. Incorrect: While it is correct that the LCV's INOPERABILITY renders T/S LCO 3.7.5 NOT MET, it is not correct that the AC exhaust fan serves a safety related purpose and as such is required by the T/S. It is plausible to believe that it does so as its name implies such (i.e. the AC emergency exhaust fans [emphasis added]). C. Incorrect: While it is correct that in accordance with the AB HVAC system description, the TDAFWP room's DC emergency exhaust fan is safety-related, it is neither correct that the AC emergency exhaust fan is required to serve a safety related function and as such is required OPERABLE by the T/S nor is it correct that the loss of a single LCV would allow the train of AFW to remain OPERABLE. It is plausible to believe such as one could reason that as long as a MDAFWP could feed one S/G that it would remain OPERABLE. D. Correct: It is correct that both the LCV and the DC emergency exhaust fan are required OPERABLE for T/S LCO 3.7.5 to be MET.
    - Unit 1 is at 100% power.
Question Number: 88  Tier:  2 Group:  1 K/A: 061 Auxiliary / Emergency Feedwater System 2.2 Equipment Control 2.2.22 Knowledge of limiting conditions for operations and safety limits. Importance Rating: 4.0  4.7
    - The following equipment is INOPERABLE:
: 1. The TDAFWP rooms DC emergency exhaust fan.
: 2. The TDAFWP rooms AC emergency exhaust fan.
: 3. 1-LCV-3-164, MDAFWP SG 1 SUPPLY.
Of the THREE equipment items listed above, which ONE of the following lists the items that are required to be OPERABLE in accordance with T/S LCO 3.7.5, AFW System?
A.           ONLY 3 B.           ONLY 2 and 3 C.           ONLY 1 and 2 D.           ONLY 1 and 3
 
CORRECT ANSWER:                                                             D DISTRACTOR ANALYSIS:
A. Incorrect: As seen in WBN-SDD-N3-30AB-4001, Auxiliary Building Heating, Ventilation, Air Conditioning System, the TDAFW pump rooms are normally ventilated by the AB air exhaust system. Two 100% emergency exhaust fans, one AC operated and one DC operated, are provided in each TDAFW pump roomThe DC-operated fan is installed to provide the required cooling in the event of a loss of all AC power, and will automatically start upon the start of the TDAFW pump. This is the only AFW pump available during a loss of all onsite and offsite AC power. The DC fan is the only means available to maintain the temperature requirements in the room. The DC fan is therefore a primary safety-related system componentThe AC fan does not serve a safety-related function. Therefore it is correct that in accordance with the AB HVAC system description, the TDAFWP rooms DC emergency exhaust fan is safety-related.
The basis for T/S LCO 3.7.5 reflects that: This requires that the two motor driven AFW pumps be OPERABLE in two diverse paths, each supplying AFW to separate steam generators. Therefore, each MDAFWP must be able to provide two S/Gs with AFW (or stated, the MDAFWPs must be able to provide all of the S/Gs). If one of the LCVs is INOPERABLE, then that train of AFW is rendered INOPERABLE.
B. Incorrect: While it is correct that the LCVs INOPERABILITY renders T/S LCO 3.7.5 NOT MET, it is not correct that the AC exhaust fan serves a safety related purpose and as such is required by the T/S. It is plausible to believe that it does so as its name implies such (i.e. the AC emergency exhaust fans [emphasis added]).
C. Incorrect: While it is correct that in accordance with the AB HVAC system description, the TDAFWP rooms DC emergency exhaust fan is safety-related, it is neither correct that the AC emergency exhaust fan is required to serve a safety related function and as such is required OPERABLE by the T/S nor is it correct that the loss of a single LCV would allow the train of AFW to remain OPERABLE. It is plausible to believe such as one could reason that as long as a MDAFWP could feed one S/G that it would remain OPERABLE.
D. Correct: It is correct that both the LCV and the DC emergency exhaust fan are required OPERABLE for T/S LCO 3.7.5 to be MET.


10 CFR Part 55: (CFR: 41.5 / 43.2 / 45.2) 10CFR55.43.b: 10 CFR 55.43(b)(2) K/A Match: K/A is matched because the applicant is required determine which items (as provided in an equipment list) affect the OPERABILITY of the AFW system.
Question Number:        88 Tier:    2  Group:      1 K/A:    061 Auxiliary / Emergency Feedwater System 2.2 Equipment Control 2.2.22 Knowledge of limiting conditions for operations and safety limits.
Importance Rating:      4.0 4.7 10 CFR Part 55:     (CFR: 41.5 / 43.2 / 45.2) 10CFR55.43.b:       10 CFR 55.43(b)(2)
K/A Match:   K/A is matched because the applicant is required determine which items (as provided in an equipment list) affect the OPERABILITY of the AFW system.
Technical  
Technical  


==Reference:==
==Reference:==
Basis for T/S LCO 3.7.5, AFW System WBN-SDD-N3-30AB-4001, Auxiliary Building Heating, Ventilation, Air Conditioning System Proposed references to be provided: None  Learning Objective: 3-OT-SYS003B 13. DESCRIBE the following aspects of TS and TRs b. The Limiting Conditions for Operation, Applicability, and Bases. Cognitive Level:     Higher   Lower X   Question Source:     New X Modified Bank   Bank     Question History: New question for the 2015-301 NRC SRO Exam Comments:
Basis for T/S LCO 3.7.5, AFW System WBN-SDD-N3-30AB-4001, Auxiliary Building Heating, Ventilation, Air Conditioning System Proposed references to     None be provided:
AFW System B 3.7.5 BASES (continued)   (continued)    Watts Bar-Unit 1 B 3.7-27      LCO   This LCO provides assurance that the AFW System will perform its design safety function to mitigate the consequences of accidents that could result in overpressurization of the reactor coolant pressure boundary. Three independent AFW pumps in three diverse trains are required to be OPERABLE to ensure the availability of RHR capability for all events accompanied by a loss of offsite power and a single failure. This is accomplished by powering two of the pumps from independent emergency buses. The third AFW pump is powered by a different means, a steam driven turbine supplied with steam from a source that is not isolated by closure of the MSIVs. The AFW System is considered OPERABLE when the components and flow paths required to provide redundant AFW flow to the steam generators are OPERABLE. This requires that the two motor driven AFW pumps be OPERABLE in two diverse paths, each supplying AFW to separate steam generators. The turbine driven AFW pump is required to be OPERABLE with redundant steam supplies from each of two main steam lines upstream of the MSIVs, and shall be capable of supplying AFW to any of the steam generators. The piping, valves, instrumentation, and controls in the required flow paths also are required to be OPERABLE.
Learning Objective:       3-OT-SYS003B
The LCO is modified by a Note indicating that one AFW train, which includes a motor driven pump, is required to be OPERABLE in MODE 4. This is because of the reduced heat removal requirements and short period of time in MODE 4 during which the AFW is required and the insufficient steam available in MODE 4 to power the turbine driven AFW pump. APPLICABILITY In MODES 1, 2, and 3, the AFW System is required to be OPERABLE in the event that it is called upon to function when the MFW is lost. In addition, the AFW System is required to supply enough makeup water to replace the steam generator secondary inventory, lost as the unit cools to MODE 4 conditions.
: 13. DESCRIBE the following aspects of TS and TRs
: b. The Limiting Conditions for Operation, Applicability, and Bases.
Cognitive Level:
Higher Lower               X Question Source:
New                 X Modified Bank Bank Question History:         New question for the 2015-301 NRC SRO Exam Comments:
 
AFW System B 3.7.5 BASES (continued)
LCO               This LCO provides assurance that the AFW System will perform its design safety function to mitigate the consequences of accidents that could result in overpressurization of the reactor coolant pressure boundary. Three independent AFW pumps in three diverse trains are required to be OPERABLE to ensure the The MDAFWPs                availability of RHR capability for all events accompanied by a loss of offsite power must be able to            and a single failure. This is accomplished by powering two of the pumps from supply 2 S/G per            independent emergency buses. The third AFW pump is powered by a different means, a steam driven turbine supplied with steam from a source that is not MDAFWP (i.e. all            isolated by closure of the MSIVs.
four S/Gs must have an operable            The AFW System is considered OPERABLE when the components and flow feed)                      paths required to provide redundant AFW flow to the steam generators are OPERABLE. This requires that the two motor driven AFW pumps be OPERABLE in two diverse paths, each supplying AFW to separate steam generators. The turbine driven AFW pump is required to be OPERABLE with redundant steam supplies from each of two main steam lines upstream of the MSIVs, and shall be capable of supplying AFW to any of the steam generators.
The piping, valves, instrumentation, and controls in the required flow paths also are required to be OPERABLE.
The LCO is modified by a Note indicating that one AFW train, which includes a motor driven pump, is required to be OPERABLE in MODE 4. This is because of the reduced heat removal requirements and short period of time in MODE 4 during which the AFW is required and the insufficient steam available in MODE 4 to power the turbine driven AFW pump.
APPLICABILITY     In MODES 1, 2, and 3, the AFW System is required to be OPERABLE in the event that it is called upon to function when the MFW is lost. In addition, the AFW System is required to supply enough makeup water to replace the steam generator secondary inventory, lost as the unit cools to MODE 4 conditions.
In MODE 4 the AFW System may be used for heat removal via the steam generators.
In MODE 4 the AFW System may be used for heat removal via the steam generators.
In MODE 5 or 6, the steam generators are not normally used for heat removal, and the AFW System is not required.
In MODE 5 or 6, the steam generators are not normally used for heat removal, and the AFW System is not required.
(continued)
Watts Bar-Unit 1                            B 3.7-27
 
WBN System              AUXILIARY BUILDING HEATING,                WBN-SDD-N3-30AB-4001 Description            VENTILATION, AIR CONDITIONING              Rev. 0038 Document                      SYSTEM (30, 31, 44)                Page 63 of 201 2.1.1  Safety Function (continued)
B. Safe shutdown earthquake (SSE)
C. Loss of offsite power (LOOP)
D. Tornado E. Flood F. Airborne radioactive contamination Note:    The TDAFW Pump Room DC powered exhaust fan is required to function during a loss of all AC power.
The required DBEs and associated safety functions for the system are tabulated in Ref 7.2.22.
The ABGTS and ABSCE serve a primary safety function by (1) providing a secondary containment barrier maintained under negative pressure during certain postulated accidents involving airborne radioactivity except a Fuel Handling Accident, and (2) providing contaminant removal sufficient to keep radioactivity levels in the air released to the environment low enough to assure compliance with the requirements of 10CFR100 (Ref 7.5.1).
Although the ABGTS and ABSCE are available to minimize the consequences of a Fuel Handling Accident, they are not required to function in order to meet the control room and offsite dose limits of 10CRF50.67 (Ref 7.5.15) based on the use of Regulatory Guide 1.183 (Alternate Source Terms) methodology (Ref 7.5.16).
Other portions of the AB HVAC System also serve a primary safety function by maintaining acceptable environmental conditions within the building as discussed in Ref 7.2.2 for protection of ESF mechanical and electrical equipment and controls following a design basis event.
Those portions of the AB HVAC System not serving a primary safety function (See paragraph 2.1.2) perform a secondary safety function by maintaining limited structural integrity during an earthquake to prevent interactions with primary safety components which could jeopardize primary safety functions.
Mechanical devices and associated instrumentation and controls and electrical equipment which perform a primary or secondary safety function are tabulated in references 7.1.10 and 7.1.11.
2.1.2  Normal Function During normal operations the AB HVAC System shall be designed to maintain acceptable environmental conditions as discussed in Ref 7.2.2 for equipment protection, personnel access, operation, inspection, maintenance, and testing; and to limit the release of radioactivity to the environment during all weather conditions.
 
WBN System              AUXILIARY BUILDING HEATING,                WBN-SDD-N3-30AB-4001 Description          VENTILATION, AIR CONDITIONING                Rev. 0038 Document                    SYSTEM (30, 31, 44)                  Page 82 of 201 3.1.3  Auxiliary Building HVAC (continued)
: 3. Additional Equipment Building HVAC The Unit 1 additional equipment building is served by three nonsafety-related air conditioning units. One unit provides air to the spaces on EL. 729, 740.5, and 752. A second unit provides air to EL. 763 and 775. The third unit provides air to the equipment spaces on El. 786.5. Grated floor openings provide an air path for the return air back to each unit. The Unit 2 additional equipment building is served by one nonsafety air conditioning unit which provides air to El. 729 and 763. Each of the air conditioning units is designed to maintain the temperature at approximately 92&deg;F dry bulb and 73&deg;F wet bulb. Condensing water is provided by the raw cooling water system.
The Additional Equipment Buildings are outside the ABSCE boundary; therefore, they are not connected to the ABGTS ventilation exhaust.
: 4. Turbine Driven Auxiliary Feedwater (TDAFW) Pump Room Exhaust The TDAFW pump rooms are normally ventilated by the AB Air Exhaust System.
Two 100% emergency exhaust fans, one (115 volt, 60Hz) AC operated and one (115 volt) DC operated, are provided in each TDAFW pump room. Each fan is sized to provide the required air flow in the room for the volume changes method of cooling. The fans are roof ventilator type venting into the general spaces of the Auxiliary Building. The DC-operated fan is installed to provide the required cooling in the event of a loss of all AC power, and will automatically start upon the start of the TDAFW pump. This is the only AFW pump available during a loss of all onsite and offsite AC power. The DC fan is the only means available to maintain the temperature requirements in the room. The DC fan is therefore a primary safety-related system component (Ref 7.2.1). See Table 9.5 for the design parameters of the DC fan. The AC fan does not serve a safety-related function and is Seismic Category I(L)B (Ref 7.2.1). Both fans are thermostatically controlled to automatically operate at a room temperature of greater than the setpoint.
: 5. Sample Room Ventilation System The sample room is ventilated by five nonsafety lab hood exhaust fans. Three fans are located on the Unit 1 side and two fans are located on the Unit 2 side.
Air enters the sample room through doors with transfer grilles and backdraft dampers. Each hood is provided with a separate exhaust fan and HEPA filter assembly. The HEPA filters located upstream from each fan have a nominal efficiency of 99.97%. A differential pressure gauge indicates the need for filter replacement. Each hood exhaust fan discharges into the General Ventilation exhaust system.
: 6. Main Steam Valve Vault Ventilation System The Main steam valve vault rooms (south and north) each have an independent nonsafety ventilation system consisting of two roof-mounted exhaust fans. The fans draw outside air for room cooling through a wall opening near the floor.


WBN System Description Document AUXILIARY BUILDING HEATING, VENTILATION, AIR CONDITIONING SYSTEM (30, 31, 44) WBN-SDD-N3-30AB-4001 Rev. 0038 Page 63 of 201  2.1.1 Safety Function (continued)    B. Safe shutdown earthquake (SSE) C. Loss of offsite power (LOOP)
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
D. Tornado E. Flood F. Airborne radioactive contamination Note: The TDAFW Pump Room DC powered exhaust fan is required to function during a loss of all AC power. The required DBEs and associated safety functions for the system are tabulated in Ref 7.2.22. The ABGTS and ABSCE serve a primary safety function by (1) providing a secondary containment barrier maintained under negative pressure during certain postulated accidents involving airborne radioactivity except a Fuel Handling Accident, and (2) providing contaminant removal sufficient to keep radioactivity levels in the air released to the environment low enough to assure compliance with the requirements of 10CFR100 (Ref 7.5.1). Although the ABGTS and ABSCE are available to minimize the consequences of a Fuel Handling Accident, they are not required to function in order to meet the control room and offsite dose limits of 10CRF50.67 (Ref 7.5.15) based on the use of Regulatory Guide 1.183 (Alternate Source Terms) methodology (Ref 7.5.16). Other portions of the AB HVAC System also serve a primary safety function by maintaining acceptable environmental conditions within the building as discussed in Ref 7.2.2 for protection of ESF mechanical and electrical equipment and controls following a design basis event. Those portions of the AB HVAC System not serving a primary safety function (See paragraph 2.1.2) perform a secondary safety function by maintaining limited structural integrity during an earthquake to prevent interactions with primary safety components which could jeopardize primary safety functions. Mechanical devices and associated instrumentation and controls and electrical equipment which perform a primary or secondary safety function are tabulated in references 7.1.10 and 7.1.11. 2.1.2 Normal Function During normal operations the AB HVAC System shall be designed to maintain acceptable environmental conditions as discussed in Ref 7.2.2 for equipment protection, personnel access, operation, inspection, maintenance, and testing; and to limit the release of radioactivity to the environment during all weather conditions.
II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]:
WBN System Description Document AUXILIARY BUILDING HEATING, VENTILATION, AIR CONDITIONING SYSTEM (30, 31, 44) WBN-SDD-N3-30AB-4001 Rev. 0038 Page 82 of 201  3.1.3 Auxiliary Building HVAC (continued)    3. Additional Equipment Building HVAC The Unit 1 additional equipment building is served by three nonsafety-related air conditioning units. One unit provides air to the spaces on EL. 729, 740.5, and 752. A second unit provides air to EL. 763 and 775. The third unit provides air to the equipment spaces on El. 786.5. Grated floor openings provide an air path for the return air back to each unit. The Unit 2 additional equipment building is served by one nonsafety air conditioning unit which provides air to El. 729 and 763. Each of the air conditioning units is designed to maintain the temperature at approximately 92&deg;F dry bulb and 73&deg;F wet bulb. Condensing water is provided by the raw cooling water system. The Additional Equipment Buildings are outside the ABSCE boundary; therefore, they are not connected to the ABGTS ventilation exhaust. 4. Turbine Driven Auxiliary Feedwater (TDAFW) Pump Room Exhaust The TDAFW pump rooms are normally ventilated by the AB Air Exhaust System. Two 100% emergency exhaust fans, one (115 volt, 60Hz) AC operated and one (115 volt) DC operated, are provided in each TDAFW pump room. Each fan is sized to provide the required air flow in the room for the volume changes method of cooling. The fans are roof ventilator type venting into the general spaces of the Auxiliary Building. The DC-operated fan is installed to provide the required cooling in the event of a loss of all AC power, and will automatically start upon the start of the TDAFW pump. This is the only AFW pump available during a loss of all onsite and offsite AC power. The DC fan is the only means available to maintain the temperature requirements in the room. The DC fan is therefore a primary safety-related system component (Ref 7.2.1). See Table 9.5 for the design parameters of the DC fan. The AC fan does not serve a safety-related function and is Seismic Category I(L)B (Ref 7.2.1). Both fans are thermostatically controlled to automatically operate at a room temperature of greater than the setpoint. 5. Sample Room Ventilation System The sample room is ventilated by five nonsafety lab hood exhaust fans. Three fans are located on the Unit 1 side and two fans are located on the Unit 2 side. Air enters the sample room through doors with transfer grilles and backdraft dampers. Each hood is provided with a separate exhaust fan and HEPA filter assembly. The HEPA filters located upstream from each fan have a nominal efficiency of 99.97%. A differential pressure gauge indicates the need for filter replacement. Each hood exhaust fan discharges into the General Ventilation exhaust system. 6. Main Steam Valve Vault Ventilation System The Main steam valve vault rooms (south and north) each have an independent nonsafety ventilation system consisting of two roof-mounted exhaust fans. The fans draw outside air for room cooling through a wall opening near the floor.
A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)]
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 3 of 16  II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]: A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)] Some examples of SRO exam items for this topic include:
Some examples of SRO exam items for this topic include:
* Reporting requirements when the maximum licensed thermal power output is exceeded.
* Reporting requirements when the maximum licensed thermal power output is exceeded.
* Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems, fire doors, etc.
* Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems, fire doors, etc.
* The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
* The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
* National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
* National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
* Processes for TS and FSAR changes. Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic. B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)] Some examples of SRO exam items for this topic include:
* Processes for TS and FSAR changes.
Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic.
B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]
Some examples of SRO exam items for this topic include:
* Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
* Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
* Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
* Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
* Knowledge of TS bases that are required to analyze TS required actions and terminology.
* Knowledge of TS bases that are required to analyze TS required actions and terminology.
* Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).
* Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of  1 hour action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.  
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of  1 hour action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. ROs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Page 3 of 16


SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. RO's are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 4 of 16 ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)         Above this line RO knowledge Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 5 of 16 Figure 1:  Screening for SRO-only linked to 10 CFR 55.43(b)(2) (Tech Specs)       
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)
RO knowledge Above this line Page 4 of 16


Can question be answered solely by knowing  1 hour TS/TRM Action? RO question YesNoCan question be answered solely by knowing the LCO/TRM information listed "above-the-line?" YesRO question NoCan question be answered solely by knowing the TS Safety Limits? YesRO question NoDoes the question involve one or more of the following for TS, TRM, or ODCM?
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)
(Tech Specs)
Can question be answered solely by knowing  1       Yes hour TS/TRM Action?                                                 RO question No Can question be answered solely by knowing the       Yes LCO/TRM information listed above-the-line?                         RO question No Can question be answered solely by knowing the       Yes TS Safety Limits?                                                   RO question The question requires  that the applicant consider both the information in the T/S LCO basis as well No                                        as that contained in the Aux Bldg ventilation system description to determine the compliance with T/S LCO 3.7.5.
Does the question involve one or more of the following for TS, TRM, or ODCM?
* Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
* Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)
* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)                                   Yes          SRO-only
* Knowledge of TS bases that is required to analyze TS required actions and terminology SRO-only question YesNoQuestion might not be linked to 10 CFR 55.43(b)(2) for SRO-only
* Knowledge of TS bases that is required to analyze TS                             question required actions and terminology No Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only Page 5 of 16
: 89. Given the following conditions: - Unit 1 is at 100%. - A buzzer is heard from behind the Shift Manager's desk.  
 
- The following is observed: - The crew enters 0-AOI-13, section 3.2, Loss of ERCW Pump. Which ONE of the following describes T/S LCO 3.7.8, ERCW? In accordance with T/S LCO 3.7.8, ________. A. a train of ERCW is INOPERABLE and will ONLY be restored OPERABLE when the failed pump is repaired and retested B. the failure does NOT impact the operability of the ERCW system and in accordance with OPDP-8, the crew will enter a TRACKING ONLY LCO for the ERCW system C. a train of ERCW is INOPERABLE and will be restored OPERABLE IMMEDIATELY after the crew performs step 1 of 0-AOI-13, START redundant trained ERCW Pump D. a train of ERCW is INOPERABLE and will be restored OPERABLE IMMEDIATELY after the crew performs step 4 of 0-AOI-13, ENSURE applicable emergency power selector switch selected away from failed pump CORRECT ANSWER:D DISTRACTOR ANALYSIS:  A. Incorrect, T/S LCO 3.7.8, ERCW states:  "Two ERCW trains shall be OPERABLE."  The basis for this T/S declares:  "An ERCW train is considered OPERABLE during MODES 1, 2, 3, and 4 when: Two pumps, aligned to separate shutdown boards, are OPERABLE."  This indicates that one ERCW pump must be available per shutdown board. The basis statement "Two pumps per train are aligned to receive power from different diesel generators" indicates the T/S impact of the emergency power selector switch. The emergency power selector switch designates which ERCW pump will automatically start either after a blackout or a safety injection. The ERCW pump not selected will not automatically start after either a blackout or SI. The one ERCW pump per Shutdown Board required OPERABLE by the T/S is therefore, that ERCW pump which is selected by the emergency power selector switch.
89.
Given the following conditions:
  -   Unit 1 is at 100%.
  -   A buzzer is heard from behind the Shift Managers desk.
  -   The following is observed:
    - The crew enters 0-AOI-13, section 3.2, Loss of ERCW Pump.
Which ONE of the following describes T/S LCO 3.7.8, ERCW?
In accordance with T/S LCO 3.7.8, ________.
A. a train of ERCW is INOPERABLE and will ONLY be restored OPERABLE when the failed pump is repaired and retested B. the failure does NOT impact the operability of the ERCW system and in accordance with OPDP-8, the crew will enter a TRACKING ONLY LCO for the ERCW system C. a train of ERCW is INOPERABLE and will be restored OPERABLE IMMEDIATELY after the crew performs step 1 of 0-AOI-13, START redundant trained ERCW Pump D. a train of ERCW is INOPERABLE and will be restored OPERABLE IMMEDIATELY after the crew performs step 4 of 0-AOI-13, ENSURE applicable emergency power selector switch selected away from failed pump


CORRECT ANSWER:                                                              D DISTRACTOR ANALYSIS:
A. Incorrect, T/S LCO 3.7.8, ERCW states: Two ERCW trains shall be OPERABLE. The basis for this T/S declares: An ERCW train is considered OPERABLE during MODES 1, 2, 3, and 4 when: Two pumps, aligned to separate shutdown boards, are OPERABLE. This indicates that one ERCW pump must be available per shutdown board. The basis statement Two pumps per train are aligned to receive power from different diesel generators indicates the T/S impact of the emergency power selector switch. The emergency power selector switch designates which ERCW pump will automatically start either after a blackout or a safety injection. The ERCW pump not selected will not automatically start after either a blackout or SI. The one ERCW pump per Shutdown Board required OPERABLE by the T/S is therefore, that ERCW pump which is selected by the emergency power selector switch.
In the conditions depicted in the stem of the question, the A-A ERCW pump has tripped. The buzzer, white and green indicating lights and lack of pressure indication all relate that the pump had been running and is now tripped. The emergency power selector switch can be seen in the A-A pump position. Therefore, an inoperable pump is selected by the selector switch. Therefore, it is true that T/S LCO 3.7.8 is NOT MET at the time that the indications are beheld.
In the conditions depicted in the stem of the question, the A-A ERCW pump has tripped. The buzzer, white and green indicating lights and lack of pressure indication all relate that the pump had been running and is now tripped. The emergency power selector switch can be seen in the A-A pump position. Therefore, an inoperable pump is selected by the selector switch. Therefore, it is true that T/S LCO 3.7.8 is NOT MET at the time that the indications are beheld.
It is incorrect to believe that the ERCW train will only be restored OPERABLE when the A-A ERCW pump is repaired and retested. It is plausible to believe this as one may believe that all four ERCW pumps per train are required operable by the T/S. B. Incorrect, This distractor is incorrect as detailed above. It would be correct if the emergency power selector switch had been seen in the B-A position. The B-A pump is OPERABLE and would be selected for emergency start. Therefore, the crew would enter a "tracking-only" T/S for LCO 3.7.8. C. Incorrect, This distractor is incorrect as detailed above. It is plausible given that if the question of OPERABILITY were treated in the same manner that a system such as CCS was to be, then the applicant would arrive at this distractor. Specifically, take the example of the Unit 1, "A" train CCS. If the 1B CCS pump were running (aligned as normal to the "A" train) and then tripped with the 1A pump failing to start in automatic, T/S LCO 3.7.7, CCS would at that point be NOT MET. Subsequently, if an operator took the 1A CCS pump to start and did start the pump, then the T/S LCO 3.7.7 would be MET. D. Correct, As detailed above, the only action required to restore the operability of the "A" train ERCW is to reposition the emergency power selector switch away from the failed pump.
It is incorrect to believe that the ERCW train will only be restored OPERABLE when the A-A ERCW pump is repaired and retested. It is plausible to believe this as one may believe that all four ERCW pumps per train are required operable by the T/S.
Question Number: 89  Tier:  2 Group:  1 K/A: 076 Service Water System (SWS) A2 Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based on those predictions, use procedures to Correct: control, or mitigate the consequences of those malfunctions or operations:
B. Incorrect, This distractor is incorrect as detailed above. It would be correct if the emergency power selector switch had been seen in the B-A position.
A2.01 Loss of SWS Importance Rating: 3.5  3.7  10 CFR Part 55: (CFR: 41.5 / 43.5 / 45/3 / 45/13)
The B-A pump is OPERABLE and would be selected for emergency start. Therefore, the crew would enter a tracking-only T/S for LCO 3.7.8.
C. Incorrect, This distractor is incorrect as detailed above. It is plausible given that if the question of OPERABILITY were treated in the same manner that a system such as CCS was to be, then the applicant would arrive at this distractor. Specifically, take the example of the Unit 1, A train CCS. If the 1B CCS pump were running (aligned as normal to the A train) and then tripped with the 1A pump failing to start in automatic, T/S LCO 3.7.7, CCS would at that point be NOT MET. Subsequently, if an operator took the 1A CCS pump to start and did start the pump, then the T/S LCO 3.7.7 would be MET.
D. Correct, As detailed above, the only action required to restore the operability of the A train ERCW is to reposition the emergency power selector switch away from the failed pump.


10CFR55.43.b: 10 CFR 55.43(b)(2) K/A Match: The K/A is matched because given an entry into 0-AOI-13, Loss of ERCW (Loss of SWS) the applicant must predict the impact on the OPERABILITY of ERCW (SWS) which is had. Subsequently, the applicant must use procedures (0-AOI-13 and the T/S) to mitigate the impact that the Loss of ERCW had upon the OPERABILITY of the system. Technical  
Question Number:        89 Tier:    2  Group:        1 K/A:    076 Service Water System (SWS)
A2 Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based on those predictions, use procedures to Correct: control, or mitigate the consequences of those malfunctions or operations:
A2.01 Loss of SWS Importance Rating:      3.5 3.7 10 CFR Part 55:        (CFR: 41.5 / 43.5 / 45/3 / 45/13) 10CFR55.43.b:         10 CFR 55.43(b)(2)
K/A Match:   The K/A is matched because given an entry into 0-AOI-13, Loss of ERCW (Loss of SWS) the applicant must predict the impact on the OPERABILITY of ERCW (SWS) which is had. Subsequently, the applicant must use procedures (0-AOI-13 and the T/S) to mitigate the impact that the Loss of ERCW had upon the OPERABILITY of the system.
1-45W760-67-1 Technical  


==Reference:==
==Reference:==
1-45W760-67-1 0-AOI-13, Loss of Essential Raw Cooling Water T/S LCO 3.7.8, ERCW Proposed references to be provided: None  Learning Objective: 3-OT-SYS067A 12. DESCRIBE the following aspects of TS and TRs
0-AOI-13, Loss of Essential Raw Cooling Water T/S LCO 3.7.8, ERCW Proposed references to     None be provided:
: b. The Limiting Conditions for Operation, Applicability, and Bases. Cognitive Level:     Higher X Lower     Question Source:     New X Modified Bank   Bank     Question History: New question for the 2015-301 NRC SRO Exam Comments:
Learning Objective:         3-OT-SYS067A
ERCW B 3.7.8      (continued)    Watts Bar-Unit 1 B 3.7-43        B 3.7  PLANT SYSTEMS B 3.7.8  Essential Raw Cooling Water (ERCW) System
: 12. DESCRIBE the following aspects of TS and TRs
: b. The Limiting Conditions for Operation, Applicability, and Bases.
Cognitive Level:
Higher               X Lower Question Source:
New                 X Modified Bank Bank Question History:
New question for the 2015-301 NRC SRO Exam Comments:


BASES BACKGROUND The ERCW provides a heat sink for the removal of process and operating heat from safety related components during a Design Basis Accident (DBA) or transient. During normal operation, and a normal shutdown, the ERCW System also provides this function for various safety related and nonsafety related components. The safety related function is covered by this LCO. The shared ERCW system consists of eight 50% ERCW pumps, four traveling water screens, four screen wash pumps, four strainers, associated piping, valves, and instrumentation. Water for the ERCW system enters two separate sump areas of the pumping station through four traveling water screens, two for each sump. Four ERCW pumping units, all on the same plant train, take suction from one of the sumps, and four more on the opposite plant train take suction from the other sump. One set of pumps and associated equipment is designated Train A, and the other Train B. These trains are redundant and are normally maintained separate and independent of each other. Each set of four pumps discharges into a common manifold, from which two separate headers (1A and 2A for Train A, and 1B and 2B for Train B) each with its own automatic backwashing strainer, supply water to the various system users. Two pumps per train are adequate to supply worst case conditions. Two pumps per train are aligned to receive power from different diesel generators. Operator designated pumps and valves are remote and manually aligned, except in the unlikely event of a loss-of-coolant accident (LOCA). The pumps are automatically started upon receipt of a safety injection (SI) signal, and some essential valves are aligned to their post-accident positions. Some manual realignments of motor-operated valves (MOVs) are necessary. The ERCW System also provides emergency makeup to the Component Cooling System (CCS) and is the backup water supply to the Auxiliary Feedwater System.       Additional information about the design and operation of the ERCW, along with a list of the components served, is presented in the FSAR, Section 9.2.1  
ERCW B 3.7.8 B 3.7 PLANT SYSTEMS B 3.7.8 Essential Raw Cooling Water (ERCW) System BASES BACKGROUND             The ERCW provides a heat sink for the removal of process and operating heat from safety related components during a Design Basis Accident (DBA) or transient. During normal operation, and a normal shutdown, the ERCW System also provides this function for various safety related and nonsafety related components. The safety related function is covered by this LCO.
The shared ERCW system consists of eight 50% ERCW pumps, four traveling water screens, four screen wash pumps, four strainers, associated piping, valves, and instrumentation.
Water for the ERCW system enters two separate sump areas of the pumping station through four traveling water screens, two for each sump. Four ERCW pumping units, all on the same plant train, take suction from one of the sumps, The ERCW pumps                      and four more on the opposite plant train take suction from the other sump. One set of pumps and associated equipment is designated Train A, and the other which will start                    Train B. These trains are redundant and are normally maintained separate and upon the receipt of                independent of each other. Each set of four pumps discharges into a common the SI signal are                  manifold, from which two separate headers (1A and 2A for Train A, and 1B and those selected by                  2B for Train B) each with its own automatic backwashing strainer, supply water to the ERCW PMP                        the various system users. Two pumps per train are adequate to supply worst case conditions. Two pumps per train are aligned to receive power from different DG POWER SEL                        diesel generators. Operator designated pumps and valves are remote and switches.                          manually aligned, except in the unlikely event of a loss-of-coolant accident (LOCA). The pumps are automatically started upon receipt of a safety injection (SI) signal, and some essential valves are aligned to their post-accident positions.
Some manual realignments of motor-operated valves (MOVs) are necessary.
The ERCW System also provides emergency makeup to the Component Cooling System (CCS) and is the backup water supply to the Auxiliary Feedwater System.
Additional information about the design and operation of the ERCW, along with a list of the components served, is presented in the FSAR, Section 9.2.1 (continued)
Watts Bar-Unit 1                                B 3.7-43


ERCW B 3.7.8 BASES   (continued)    Watts Bar-Unit 1 B 3.7-44        BACKGROUND (Ref. 1). The principal safety related function of the ERCW System is the   (continued) removal of decay heat from the reactor via the CCS. APPLICABLE The design basis of the ERCW System is for one ERCW train, in conjunction SAFETY ANALYSES with the CCS and a 100% capacity Containment Spray System and Residual   Heat Removal (RHR), to remove core decay heat following a design basis LOCA as discussed in the FSAR, Section 9.2.1 (Ref. 1). This prevents the containment sump fluid from increasing in temperature during the recirculation phase following a LOCA and provides for a gradual reduction in the temperature of this fluid as it is supplied to the Reactor Coolant System by the ECCS pumps. The ERCW System is designed to perform its function with a single failure of any active component, assuming the loss of offsite power. The ERCW System, in conjunction with the CCS, also cools the unit from RHR, as discussed in the FSAR, Section 5.5.7, (Ref. 2) entry conditions to MODE 5 during normal and post accident operations. The time required for this evolution is a function of the number of CCS and RHR System trains that are operating. One ERCW train is sufficient to remove decay heat during subsequent operations in MODES 5 and 6. This assumes a maximum ERCW temperature of 85F occurring simultaneously with maximum heat loads on the system. The ERCW System satisfies Criterion 3 of the NRC Policy Statement. LCO   Two ERCW trains are required to be OPERABLE to provide the required redundancy to ensure that the system functions to remove post accident heat loads, assuming that the worst case single active failure occurs coincident with the loss of offsite power. An ERCW train is considered OPERABLE during MODES 1, 2, 3, and 4 when:
ERCW B 3.7.8 BASES BACKGROUND       (Ref. 1). The principal safety related function of the ERCW System is the (continued)     removal of decay heat from the reactor via the CCS.
ERCW B 3.7.8  BASES    (continued)   Watts Bar-Unit 1 B 3.7-45        LCO  a. Two pumps, aligned to separate shutdown boards, are OPERABLE; and  (continued)    b. The associated piping, valves, heat exchanger, and instrumentation and  controls required to perform the safety related function are OPERABLE. APPLICABILITY In MODES 1, 2, 3, and 4, the ERCW System is a normally operating system that is required to support the OPERABILITY of the equipment serviced by the ERCW System and required to be OPERABLE in these MODES. In MODES 5 and 6, the OPERABILITY requirements of the ERCW System are determined by the systems it supports. ACTIONS  A.1    If one ERCW train is inoperable, action must be taken to restore OPERABLE status within 72 hours. In this Condition, the remaining OPERABLE ERCW train is adequate to perform the heat removal function. However, the overall reliability is reduced because a single failure in the OPERABLE ERCW train could result in loss of ERCW System function. Required Action A.1 is modified by two Notes. The first Note indicates that the applicable Conditions and Required Actions of LCO 3.8.1, "AC Sources Operating," should be entered if an inoperable ERCW train results in an inoperable emergency diesel generator. The second Note indicates that the applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops MODE 4," should be entered if an inoperable ERCW train results in an inoperable decay heat removal train. This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components. The 72 hour Completion Time is based on the redundant capabilities afforded by the OPERABLE train, and the low probability of a DBA occurring during this time period.      B.1 and B.2    If the ERCW train cannot be restored to OPERABLE status within the associated Completion Time, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least   
APPLICABLE       The design basis of the ERCW System is for one ERCW train, in conjunction SAFETY ANALYSES with the CCS and a 100% capacity Containment Spray System and Residual Heat Removal (RHR), to remove core decay heat following a design basis LOCA as discussed in the FSAR, Section 9.2.1 (Ref. 1). This prevents the containment sump fluid from increasing in temperature during the recirculation phase following a LOCA and provides for a gradual reduction in the temperature of this fluid as it is supplied to the Reactor Coolant System by the ECCS pumps. The ERCW System is designed to perform its function with a single failure of any active component, assuming the loss of offsite power.
The ERCW System, in conjunction with the CCS, also cools the unit from RHR, as discussed in the FSAR, Section 5.5.7, (Ref. 2) entry conditions to MODE 5 during normal and post accident operations. The time required for this evolution is a function of the number of CCS and RHR System trains that are operating.
One ERCW train is sufficient to remove decay heat during subsequent operations in MODES 5 and 6. This assumes a maximum ERCW temperature of 85 F occurring simultaneously with maximum heat loads on the system.
The ERCW System satisfies Criterion 3 of the NRC Policy Statement.
LCO             Two ERCW trains are required to be OPERABLE to provide the required redundancy to ensure that the system functions to remove post accident heat loads, assuming that the worst case single active failure occurs coincident with the loss of offsite power.
An ERCW train is considered OPERABLE during MODES 1, 2, 3, and 4 when:
(continued)
Watts Bar-Unit 1                           B 3.7-44


WBN Unit 0 Loss of Essential Raw Cooling Water 0-AOI-13 Rev. 0001 Page 5 of 88  3.0 OPERATOR ACTIONS 3.1 Diagnostics IF GO TO SECTION PAGE Loss of ERCW pump or indications of broken pump shaft: Motor trip out alarm OR Low amps and discharge pressure on running pump 3.2 6 Supply Header Rupture in Auxiliary Building; HIGH flow on supply header AND Building flood alarm LIT. 3.3 8 Supply Header Rupture in Yard/Downstream of Strainer: Strainer DP alarm LIT AND LOW flow on individual supply header with LOW pressure on IPS supply header.
The aligned to a SDBD means that ERCW the ERCW PMP B 3.7.8 DG POWER SEL BASESswitch is positioned to the pump.
If IPS strainer room sump alarm is LIT rupture may be downstream of strainer in strainer room. 3.4 17 Plugged Strainer: Strainer DP alarm LIT AND LOW flow on individual supply header with HIGH pressure indicated on IPS supply header. 3.4 17 Supply Header Rupture in IPS; Supply headers flow LOW AND IPS header pressure LOW with Strainer DP alarm DARK, AND IPS strainer room sump alarm LIT. 3.5 28 Discharge Header Rupture in Auxiliary Building: Building flood alarm LIT AND Supply header flows NORMAL. 3.6 36 Loss of flow on ALL ERCW supply headers 3.7 42 WBN Unit 0 Loss of Essential Raw Cooling Water 0-AOI-13 Rev. 0001    Page 6 of 88  Step Action/Expected Response Response Not Obtained 3.2 Loss of ERCW Pump 1. CHECK header pressure and flows adequate for current conditions. START redundant trained ERCW Pump. 2. ENSURE pump amps NORMAL. 3. PLACE failed pump HS in PULL TO LOCK. 4. ENSURE applicable emergency power selector switch selected away from failed pump. 5. DISPATCH personnel to determine reason for pump failure. 6. ENSURE header pressures and flows return to expected values for existing plant conditions. IF ERCW header pressures and flows cannot be returned to NORMAL, THEN **GO TO Section 3.1 Diagnostics to evaluate for a potential rupture.
LCO                  a.      Two pumps, aligned to separate shutdown boards, are OPERABLE; and (continued)
WBN Unit 0 Loss of Essential Raw Cooling Water 0-AOI-13 Rev. 0001    3.2 Loss of ERCW Pump (continued)   Page 7 of 88  Step Action/Expected Response Response Not Obtained  7. CLOSE discharge valve on failed pump. A TRAIN PUMPS DISCHARGE VALVE B TRAIN PUMPS DISCHARGE VALVEA-A B-A C-A D-A 0-ISV-67-504A 0-ISV-67-504B 0-ISV-67-504C 0-ISV-67-504D E-B F-B G-B H-B 0-ISV-67-504E 0-ISV-67-504F 0-ISV-67-504G 0-ISV-67-504H 8. INITIATE repair. 9. REFER TO Tech Spec 3.7.8, Essential Raw Cooling Water System (ERCW). 10. RETURN TO Instruction in effect. End of Section Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 3 of 16  II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]: A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)] Some examples of SRO exam items for this topic include:
: b.      The associated piping, valves, heat exchanger, and instrumentation and controls required to perform the safety related function are OPERABLE.
APPLICABILITY        In MODES 1, 2, 3, and 4, the ERCW System is a normally operating system that is required to support the OPERABILITY of the equipment serviced by the ERCW System and required to be OPERABLE in these MODES.
In MODES 5 and 6, the OPERABILITY requirements of the ERCW System are determined by the systems it supports.
ACTIONS              A.1 If one ERCW train is inoperable, action must be taken to restore OPERABLE status within 72 hours. In this Condition, the remaining OPERABLE ERCW train is adequate to perform the heat removal function. However, the overall reliability is reduced because a single failure in the OPERABLE ERCW train could result in loss of ERCW System function. Required Action A.1 is modified by two Notes.
The first Note indicates that the applicable Conditions and Required Actions of LCO 3.8.1, "AC Sources Operating," should be entered if an inoperable ERCW train results in an inoperable emergency diesel generator. The second Note indicates that the applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops MODE 4," should be entered if an inoperable ERCW train results in an inoperable decay heat removal train. This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components. The 72 hour Completion Time is based on the redundant capabilities afforded by the OPERABLE train, and the low probability of a DBA occurring during this time period.
B.1 and B.2 If the ERCW train cannot be restored to OPERABLE status within the associated Completion Time, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least (continued)
Watts Bar-Unit 1                              B 3.7-45
 
WBN           Loss of Essential Raw Cooling Water   0-AOI-13 Unit 0                                                Rev. 0001 3.0     OPERATOR ACTIONS 3.1     Diagnostics IF                                                           GO TO PAGE SECTION Loss of ERCW pump or indications of broken pump shaft:         3.2    6 Motor trip out alarm OR Low amps and discharge pressure on running pump Supply Header Rupture in Auxiliary Building;                   3.3    8 HIGH flow on supply header AND Building flood alarm LIT.
Supply Header Rupture in Yard/Downstream of Strainer:         3.4    17 Strainer DP alarm LIT AND LOW flow on individual supply header with LOW pressure on IPS supply header.
If IPS strainer room sump alarm is LIT rupture may be downstream of strainer in strainer room.
Plugged Strainer:                                              3.4   17 Strainer DP alarm LIT AND LOW flow on individual supply header with HIGH pressure indicated on IPS supply header.
Supply Header Rupture in IPS;                                 3.5    28 Supply headers flow LOW AND IPS header pressure LOW with Strainer DP alarm DARK, AND IPS strainer room sump alarm LIT.
Discharge Header Rupture in Auxiliary Building:               3.6    36 Building flood alarm LIT AND Supply header flows NORMAL.
Loss of flow on ALL ERCW supply headers                       3.7   42 Page 5 of 88
 
WBN           Loss of Essential Raw Cooling Water     0-AOI-13 Unit 0                                                Rev. 0001 Step   Action/Expected Response                 Response Not Obtained 3.2     Loss of ERCW Pump
: 1.     CHECK header pressure and flows         START redundant trained ERCW adequate for current conditions.         Pump.
: 2.     ENSURE pump amps NORMAL.
: 3.     PLACE failed pump HS in PULL TO LOCK.
Notice that T/S
: 4.     ENSURE applicable emergency                               LCO 3.7.8 is NOT power selector switch selected away                       met from the point from failed pump.                                         of pump loss until the completion of step 4.
: 5.     DISPATCH personnel to determine reason for pump failure.
: 6.     ENSURE header pressures and flows        IF ERCW header pressures and flows return to expected values for existing   cannot be returned to NORMAL, THEN plant conditions.                        **GO TO Section 3.1 Diagnostics to evaluate for a potential rupture.
Page 6 of 88
 
WBN         Loss of Essential Raw Cooling Water     0-AOI-13 Unit 0                                                Rev. 0001 Step   Action/Expected Response                Response Not Obtained 3.2     Loss of ERCW Pump (continued)
: 7.     CLOSE discharge valve on failed pump.
A TRAIN     DISCHARGE VALVE           B TRAIN     DISCHARGE VALVE PUMPS                                  PUMPS A-A         0-ISV-67-504A             E-B        0-ISV-67-504E B-A          0-ISV-67-504B            F-B        0-ISV-67-504F C-A          0-ISV-67-504C            G-B        0-ISV-67-504G D-A          0-ISV-67-504D            H-B        0-ISV-67-504H
: 8.     INITIATE repair.
: 9.     REFER TO Tech Spec 3.7.8,                                 Again, T/S LCO Essential Raw Cooling Water System                       3.7.8 was not met (ERCW).                                                   for the mentioned period. After such,
: 10. RETURN TO Instruction in effect.                         a TRACKING ONLY LCO is End of Section entered.
Page 7 of 88
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]:
A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)]
Some examples of SRO exam items for this topic include:
* Reporting requirements when the maximum licensed thermal power output is exceeded.
* Reporting requirements when the maximum licensed thermal power output is exceeded.
* Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems, fire doors, etc.
* Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems, fire doors, etc.
* The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
* The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
* National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
* National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
* Processes for TS and FSAR changes. Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic. B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)] Some examples of SRO exam items for this topic include:
* Processes for TS and FSAR changes.
Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic.
B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]
Some examples of SRO exam items for this topic include:
* Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
* Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
* Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
* Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
* Knowledge of TS bases that are required to analyze TS required actions and terminology.
* Knowledge of TS bases that are required to analyze TS required actions and terminology.
* Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).
* Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of  1 hour action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.  
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of  1 hour action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. ROs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Page 3 of 16


SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. RO's are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 4 of 16 ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)         Above this line RO knowledge Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 5 of 16 Figure 1:  Screening for SRO-only linked to 10 CFR 55.43(b)(2) (Tech Specs)       
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)
RO knowledge Above this line Page 4 of 16


Can question be answered solely by knowing  1 hour TS/TRM Action? RO question YesNoCan question be answered solely by knowing the LCO/TRM information listed "above-the-line?" YesRO question NoCan question be answered solely by knowing the TS Safety Limits? YesRO question NoDoes the question involve one or more of the following for TS, TRM, or ODCM?
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)
(Tech Specs)
Can question be answered solely by knowing  1       Yes hour TS/TRM Action?                                           RO question No Can question be answered solely by knowing the       Yes LCO/TRM information listed above-the-line?                   RO question No Can question be answered solely by knowing the       Yes TS Safety Limits?                                             RO question No Does the question involve one or more of the following for TS, TRM, or ODCM?
* Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
* Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)
* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)                             Yes      SRO-only
* Knowledge of TS bases that is required to analyze TS required actions and terminology SRO-only question YesNoQuestion might not be linked to 10 CFR 55.43(b)(2) for SRO-only
* Knowledge of TS bases that is required to analyze TS                   question required actions and terminology No Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only Page 5 of 16
: 90. Given the following conditions: - Unit 1 is at 100% power. - Containment Pressure Transmitter 1-PDT-30-43 (Channel III) FAILED and is out of service with the channel bistables positioned as required by Tech Specs. - The Surveillance Instruction for 1-PDT-30-44 (Channel II) is NOW due. Which ONE of the following describes the required action for performing the Surveillance Instruction on 1-PDT-30-44 and the impact on Containment Spray actuation? 1-PDT-30-43 is required to be placed in the ____(1)____ position AND subsequent testing of 1-PDT-30-44 will ____(2)____ a valid AUTOMATIC Containment Spray actuation.
 
A. (1) BYPASS (2) ALLOW B. (1) BYPASS (2) PREVENT C. (1) TRIPPED (2) ALLOW D. (1) TRIPPED (2) PREVENT CORRECT ANSWER:A DISTRACTOR ANALYSIS:   A. Correct, 1-PDT-30-43 would be placed in the bypass position as identified in the Tech Spec 3.3.2 Bases. The Required Action for LCO 3.3.2 Condition E has a Note that allows a channel to be bypassed for up to 12 hours for surveillance testing. This Note is explained in the Tech Spec Bases. There are 4 channels provided for the Hi-Hi containment function to actuate and it takes 2 of the 4 to generate the signal (and 2 channels remain in service). B. Incorrect, 1-PDT-30-43 would be placed in the bypass position but the testing of 1-PDT-30-44 will not prevent a valid containment spray actuation from occurring even though the HI-HI bistables would be tested in the bypass position. C. Incorrect, 1-PDT-30-43 will not be placed to the trip position and subsequent testing of 1-PDT-30-44 will still allow valid automatic actuation of the containment spray during the surveillance test because the other 2 channels still provide for the protection of the function. Plausible because the positioning of bistables to the trip position is a normal process performed to meet Tech Spec requirements to ensure a function can be met and in this COT the Hi containment pressure SI bistable for this instrument would be placed in the Trip position unless another channel was to be tested. D. Incorrect, 1-PDT-30-43 will not be placed to the trip position and subsequent testing of 1-PDT-30-44 will not prevent a valid automatic actuation of the containment spray during the surveillance test because the other 2 channels still provide for the protection of the function. Plausible because the positioning of bistables to the trip position is a normal process performed to meet Tech Spec requirements to ensure a function can be met (and in this COT the Hi containment pressure SI bistable for this instrument would be placed in the Trip position unless testing of another channel was required.
90.
Question Number: 90  Tier:  2 Group:  1 K/A: 026 Containment Spray System A2 Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; and (b) based on those predictions, use procedures to Correct: control, or mitigate the consequences of those malfunctions or operations:
Given the following conditions:
A2.03 Failure of ESF Importance Rating: 4.1  4.4  10 CFR Part 55: (CFR: 41.5 / 43.5 / 45.3 / 45.13)
    - Unit 1 is at 100% power.
    - Containment Pressure Transmitter 1-PDT-30-43 (Channel III) FAILED and is out of service with the channel bistables positioned as required by Tech Specs.
    - The Surveillance Instruction for 1-PDT-30-44 (Channel II) is NOW due.
Which ONE of the following describes the required action for performing the Surveillance Instruction on 1-PDT-30-44 and the impact on Containment Spray actuation?
1-PDT-30-43 is required to be placed in the ____(1)____ position AND subsequent testing of 1-PDT-30-44 will ____(2)____ a valid AUTOMATIC Containment Spray actuation.
A.     (1)   BYPASS (2)   ALLOW B.     (1)   BYPASS (2)   PREVENT C.     (1)   TRIPPED (2)   ALLOW D.     (1)   TRIPPED (2)   PREVENT
 
CORRECT ANSWER:                                                                   A DISTRACTOR ANALYSIS:
A. Correct, 1-PDT-30-43 would be placed in the bypass position as identified in the Tech Spec 3.3.2 Bases. The Required Action for LCO 3.3.2 Condition E has a Note that allows a channel to be bypassed for up to 12 hours for surveillance testing. This Note is explained in the Tech Spec Bases. There are 4 channels provided for the Hi-Hi containment function to actuate and it takes 2 of the 4 to generate the signal (and 2 channels remain in service).
B. Incorrect, 1-PDT-30-43 would be placed in the bypass position but the testing of 1-PDT-30-44 will not prevent a valid containment spray actuation from occurring even though the HI-HI bistables would be tested in the bypass position.
C. Incorrect, 1-PDT-30-43 will not be placed to the trip position and subsequent testing of 1-PDT-30-44 will still allow valid automatic actuation of the containment spray during the surveillance test because the other 2 channels still provide for the protection of the function. Plausible because the positioning of bistables to the trip position is a normal process performed to meet Tech Spec requirements to ensure a function can be met and in this COT the Hi containment pressure SI bistable for this instrument would be placed in the Trip position unless another channel was to be tested.
D. Incorrect, 1-PDT-30-43 will not be placed to the trip position and subsequent testing of 1-PDT-30-44 will not prevent a valid automatic actuation of the containment spray during the surveillance test because the other 2 channels still provide for the protection of the function. Plausible because the positioning of bistables to the trip position is a normal process performed to meet Tech Spec requirements to ensure a function can be met (and in this COT the Hi containment pressure SI bistable for this instrument would be placed in the Trip position unless testing of another channel was required.


10CFR55.43.b: 10 CFR 55.43(b)(2) K/A Match: Applicant must determine how the bistables will be concurrently configured on a failed Containment Spray System actuation ESF transmitter and an ESF transmitter which is required to be tested in order to run the surveillance instruction and how the function is maintained as identified in the Technical Specification bases.
Question Number:        90 Tier:    2  Group:        1 K/A:    026 Containment Spray System A2 Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; and (b) based on those predictions, use procedures to Correct: control, or mitigate the consequences of those malfunctions or operations:
A2.03 Failure of ESF Importance Rating:      4.1 4.4 10 CFR Part 55:        (CFR: 41.5 / 43.5 / 45.3 / 45.13) 10CFR55.43.b:         10 CFR 55.43(b)(2)
K/A Match:   Applicant must determine how the bistables will be concurrently configured on a failed Containment Spray System actuation ESF transmitter and an ESF transmitter which is required to be tested in order to run the surveillance instruction and how the function is maintained as identified in the Technical Specification bases.
Technical  
Technical  


==Reference:==
==Reference:==
T/S LCO 3.3.2, ESFAS Instrumentation T/S Basis for LCO 3.3.2 1-47W611-88-1   Proposed references to be provided: None  Learning Objective: 3-OT-SYS072A 11. DESCRIBE the following aspects of TS and TRs
T/S LCO 3.3.2, ESFAS Instrumentation T/S Basis for LCO 3.3.2 1-47W611-88-1 Proposed references to     None be provided:
: b. The Limiting Conditions for Operation, Applicability, and Bases. Cognitive Level:     Higher   Lower X   Question Source:     New   Modified Bank   Bank X   Question History: Bank question 026A2.03 88. Used on the 11/2009 WBN NRC exam. Comments:
Learning Objective:         3-OT-SYS072A
ESFAS Instrumentation 3.3.2       Watts Bar-Unit 1 3.3-24      3.3  INSTRUMENTATION 3.3.2 Engineered Safety Feature Actuation System (ESFAS) Instrumentation   LCO 3.3.2 The ESFAS instrumentation for each Function in Table 3.3.2-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.2-1. ACTIONS -------------------------------------NOTE------------------------------------- Separate Condition entry is allowed for each Function. ------------------------------------------------------------------------------  CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions with one or more required channels or trains inoperable. A.1 Enter the Condition referenced in Table 3.3.2-1 for the channel(s) or train(s). Immediately  B. One channel or train inoperable. B.1 Restore channel or train to OPERABLE status. OR B.2.1 Be in MODE 3. AND B.2.2 Be in MODE 5. 48 hours    54 hours    84 hours   (continued)
: 11. DESCRIBE the following aspects of TS and TRs
ESFAS Instrumentation 3.3.2        Watts Bar-Unit 1 3.3-26 Amendment 68    ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME E. One Containment   Pressure channel inoperable. E.1 ---------------NOTE---------------- One channel may be   bypassed for up to 12 hours for surveillance testing. ----------------------------------------  Place channel in bypass. OR E.2.1 Be in MODE 3. AND E.2.2 Be in MODE 4.         72 hours    78 hours    84 hours F. One channel or train inoperable. F.1 Restore channel or   train to OPERABLE   status. OR F.2.1 Be in MODE 3. AND F.2.2 Be in MODE 4. 48 hours      54 hours    60 hours   (continued)
: b. The Limiting Conditions for Operation, Applicability, and Bases.
ESFAS Instrumentation 3.3.2      Watts Bar-Unit 1 3.3-34      Table 3.3.2-1 (page 1 of 7) Engineered Safety Feature Actuation System Instrumentation      FUNCTION  APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS    REQUIRED CHANNELS    CONDITIONS  SURVEILLANCE REQUIREMENTS  ALLOWABLE VALUE  NOMINAL TRIP SETPOINT  1. Safety Injection        a. Manual  1, 2, 3, 4 2 B SR 3.3.2.8 NA NA  Initiation        b. Automatic 1, 2, 3, 4 2 trains C SR 3.3.2.2 NA NA  Actuation Logic    SR 3.3.2.3    and Actuation    SR 3.3.2.5    Relays    SR 3.3.2.7    c. Containment 1, 2, 3 3 D SR 3.3.2.1  1.6 psig 1.5 psig  Pressure-    SR 3.3.2.4    High    SR 3.3.2.9        SR 3.3.2.10    d. Pressurizer 1, 2, 3(a) 3 D SR 3.3.2.1  1864.8 psig 1870 psig  Pressure-Low    SR 3.3.2.4        SR 3.3.2.9        SR 3.3.2.10    e. Steam Line 1, 2, 3(a) 3 per steam D SR 3.3.2.1  666.6(b) psig 675(b) psig  Pressure-Low  line  SR 3.3.2.4        SR 3.3.2.9        SR 3.3.2.10    2. Containment Spray        a. Manual  1, 2, 3, 4 2 per train, B SR 3.3.2.8 NA NA  Initiation  2 trains      b. Automatic 1, 2, 3, 4 2 trains C SR 3.3.2.2 NA NA  Actuation Logic    SR 3.3.2.3    and Actuation    SR 3.3.2.5    Relays        c. Containment 1, 2, 3 4 E SR 3.3.2.1  2.9 psig 2.8 psig  Pressure-    SR 3.3.2.4    High High    SR 3.3.2.9        SR 3.3.2.10            (continued)  (a) Above the P-11 (Pressurizer Pressure) Interlock.
Cognitive Level:
Higher Lower               X Question Source:
New Modified Bank Bank                 X Question History:           Bank question 026A2.03 88. Used on the 11/2009 WBN NRC exam.
Comments:
 
ESFAS Instrumentation 3.3.2 3.3 INSTRUMENTATION 3.3.2 Engineered Safety Feature Actuation System (ESFAS) Instrumentation LCO 3.3.2                 The ESFAS instrumentation for each Function in Table 3.3.2-1 shall be OPERABLE.
APPLICABILITY:           According to Table 3.3.2-1.
ACTIONS
                      -------------------------------------NOTE-------------------------------------
Separate Condition entry is allowed for each Function.
CONDITION                                   REQUIRED ACTION                             COMPLETION TIME A. One or more Functions with one             A.1       Enter the Condition referenced               Immediately or more required channels or                        in Table 3.3.2-1 for the trains inoperable.                                  channel(s) or train(s).
B. One channel or train inoperable.           B.1       Restore channel or train to                   48 hours OPERABLE status.
OR B.2.1     Be in MODE 3.                                 54 hours AND B.2.2     Be in MODE 5.                                 84 hours (continued)
Watts Bar-Unit 1                                           3.3-24
 
ESFAS Instrumentation 3.3.2 ACTIONS (continued)
CONDITION                   REQUIRED ACTION                         COMPLETION TIME E. One Containment             E.1   ---------------NOTE----------------
Pressure channel inoperable.      One channel may be bypassed for up to 12 hours for surveillance testing.
Place channel in bypass.                 72 hours OR E.2.1 Be in MODE 3.                           78 hours AND E.2.2 Be in MODE 4.                           84 hours F. One channel or train       F.1   Restore channel or                       48 hours inoperable.                      train to OPERABLE status.
OR F.2.1 Be in MODE 3.                           54 hours AND F.2.2 Be in MODE 4.                           60 hours (continued)
Watts Bar-Unit 1                         3.3-26                                        Amendment 68


(b) Time constants used in the lead/lag controller are t1 50 seconds and t2  5 seconds.
ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 1 of 7)
Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES                                                                    NOMINAL OR OTHER SPECIFIED      REQUIRED                  SURVEILLANCE      ALLOWABLE          TRIP FUNCTION            CONDITIONS        CHANNELS      CONDITIONS  REQUIREMENTS        VALUE        SETPOINT
: 1. Safety Injection
: a. Manual                1, 2, 3, 4              2          B        SR 3.3.2.8          NA            NA Initiation
: b. Automatic            1, 2, 3, 4          2 trains        C        SR 3.3.2.2          NA            NA Actuation Logic                                                    SR 3.3.2.3 and Actuation                                                      SR 3.3.2.5 Relays                                                              SR 3.3.2.7
: c. Containment            1, 2, 3                3          D        SR 3.3.2.1          1.6 psig      1.5 psig Pressure-                                                          SR 3.3.2.4 High                                                                SR 3.3.2.9 SR 3.3.2.10 (a)
: d. Pressurizer            1, 2, 3                3          D        SR 3.3.2.1        1864.8 psig    1870 psig Pressure-Low                                                        SR 3.3.2.4 SR 3.3.2.9 SR 3.3.2.10    HI pressure goes
: e. Steam Line            1, 2, 3 (a) 3 per steam        D        SR 3.3.2.1 to666.6 trip.(b) psig    (b) 675 psig Pressure-Low                                line                  SR 3.3.2.4 SR 3.3.2.9 SR 3.3.2.10 HI HI pressure
: 2. Containment Spray goes to bypass
: a. Manual                1, 2, 3, 4        2 per train,      B        SR 3.3.2.8      (theNAtwo are      NA Initiation                                2 trains different).
: b. Automatic            1, 2, 3, 4          2 trains        C        SR 3.3.2.2          NA            NA Actuation Logic                                                    SR 3.3.2.3 and Actuation                                                      SR 3.3.2.5 Relays
: c. Containment            1, 2, 3                4          E        SR 3.3.2.1          2.9 psig      2.8 psig Pressure-                                                          SR 3.3.2.4 High High                                                          SR 3.3.2.9 SR 3.3.2.10 (continued)
(a)    Above the P-11 (Pressurizer Pressure) Interlock.
(b)    Time constants used in the lead/lag controller are t1   50 seconds and t2  5 seconds.
Watts Bar-Unit 1                                        3.3-34


ESFAS Instrumentation 3.3.2       Watts Bar-Unit 1 3.3-35      Table 3.3.2-1 (page 2 of 7) Engineered Safety Feature Actuation System Instrumentation     FUNCTION  APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS    REQUIRED  CHANNELS    CONDITIONS  SURVEILLANCE REQUIREMENTS   ALLOWABLE VALUE NOMINAL TRIP SETPOINT 3. Containment Isolation         a. Phase A Isolation         (1) Manual 1, 2, 3, 4 2 B SR 3.3.2.8 NA NA   Initiation         (2) Automatic 1, 2, 3, 4 2 trains C SR 3.3.2.2 NA NA   Actuation   SR 3.3.2.3     Logic and   SR 3.3.2.5     Actuation   SR 3.3.2.7     Relays         (3) Safety Refer to Function 1 (Safety Injection) for all initiation     Injection functions and requirements. b. Phase B Isolation         (1) Manual 1, 2, 3, 4 2 per train, B SR 3.3.2.8 NA NA   Initiation 2 trains       (2) Automatic 1, 2, 3, 4 2 trains C SR 3.3.2.2 NA NA   Actuation   SR 3.3.2.3     Logic and   SR 3.3.2.5     Actuation   SR 3.3.2.7     Relays         (3) Con- 1, 2, 3 4 E SR 3.3.2.1 2.9 psig 2.8 psig   tainment   SR 3.3.2.4     Pressure--   SR 3.3.2.9     High High   SR 3.3.2.10   4. Steam Line Isolation         a. Manual 1, 2(c), 3(c) 1/valve F SR 3.3.2.8 NA NA  Initiation        b. Automatic 1, 2(c), 3(c) 2 trains G SR 3.3.2.2 NA NA   Actuation     SR 3.3.2.3     Logic and   SR 3.3.2.5     Actuation         Relays       (continued)   (c) Except when all MSIVs are closed and de-activated.
ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 2 of 7)
Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES                                                                     NOMINAL OR OTHER SPECIFIED       REQUIRED                       SURVEILLANCE ALLOWABLE        TRIP FUNCTION                CONDITIONS            CHANNELS         CONDITIONS   REQUIREMENTS   VALUE       SETPOINT
: 3. Containment Isolation
: a. Phase A Isolation (1)   Manual             1, 2, 3, 4               2                 B       SR 3.3.2.8       NA           NA Initiation (2)   Automatic           1, 2, 3, 4           2 trains             C       SR 3.3.2.2       NA           NA Actuation                                                               SR 3.3.2.3 Logic and                                                               SR 3.3.2.5 Actuation                                                               SR 3.3.2.7 Relays (3)   Safety       Refer to Function 1 (Safety Injection) for all initiation Injection   functions and requirements.
: b. Phase B Isolation (1)   Manual             1, 2, 3, 4         2 per train,           B       SR 3.3.2.8       NA           NA Initiation                               2 trains (2)   Automatic           1, 2, 3, 4           2 trains             C       SR 3.3.2.2       NA           NA Actuation                                                               SR 3.3.2.3 Logic and                                                               SR 3.3.2.5 Actuation                                                               SR 3.3.2.7 Relays (3)   Con-                 1, 2, 3                 4                 E       SR 3.3.2.1     2.9 psig   2.8 psig tainment                                                               SR 3.3.2.4 Pressure--                                                             SR 3.3.2.9 High High                                                               SR 3.3.2.10
: 4. Steam Line Isolation (c)  (c)
: a. Manual                   1, 2 , 3               1/valve               F     SR 3.3.2.8       NA           NA Initiation (c)   (c)
: b. Automatic                 1, 2 , 3               2 trains             G       SR 3.3.2.2       NA           NA Actuation                                                                     SR 3.3.2.3 Logic and                                                                     SR 3.3.2.5 Actuation Relays (continued)
(c)     Except when all MSIVs are closed and de-activated.
Watts Bar-Unit 1                                            3.3-35


ESFAS Instrumentation 3.3.2       Watts Bar-Unit 1 3.3-36 Amendment 23    Table 3.3.2-1 (page 3 of 7) Engineered Safety Feature Actuation System Instrumentation   FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED  CHANNELS   CONDITIONS SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE NOMINAL TRIP SETPOINT 4. Steam Line Isolation       (continued)         c. Containment 1, 2(c), 3(c) 4 E SR 3.3.2.1 2.9 psig 2.8 psig   Pressure-   SR 3.3.2.4     High High   SR 3.3.2.9         SR 3.3.2.10     d. Steam Line         Pressure         (1) Low 1, 2(c), 3(a) (c) 3 per steam D SR 3.3.2.1 666.6(b) psig 675(b) psig     line SR 3.3.2.4         SR 3.3.2.9         SR .3.2.10     (2) Negative 3(d) (c) 3 per steam D SR 3.3.2.1 108.5(e) psi 100(e) psi   Rate-High line SR 3.3.2.4         SR 3.3.2.9         SR 3.3.2.10   5. Turbine Trip and       Feedwater Isolation         a. Automatic 1, 2(f), 3(f) 2 trains H SR 3.3.2.2 NA NA   Actuation Logic   SR 3.3.2.3     and Actuation   SR 3.3.2.5     Relays         b. SG Water 1, 2(f), 3(f) 3 per SG I SR 3.3.2.1 83.1% 82.4%   Level-High     SR 3.3.2.4     High(P-14)   SR 3.3.2.9         SR 3.3.2.10 (h)    c. Safety Refer to Function 1 (Safety Injection) for all initiation     Injection functions and requirements. d. North MSV Vault 1, 2(f), (g) 3/vault O SR 3.3.2.6 5.31 inches 4 inches   Room Water Room SR 3.3.2.9     Level - High         e. South MSV Vault 1, 2(f), (g) 3/vault O SR 3.3.2.6 4.56 inches 4 inches   Room Water Room SR 3.3.2.9     Level - High       (continued) (a) Above the P-11 (Pressurizer Pressure) interlock. (b) Time constants used in the lead/lag controller are t1 50 seconds and t2 5 seconds. (c) Except when all MSIVs are closed and de-activated. (d) Function automatically blocked above P-11 (Pressurizer Interlock) setpoint and is enabled below P-11 when safety injection on Steam Line Pressure Low is manually blocked. (e) Time constants utilized in the rate/lag controller are t3 and t4 50 seconds. (f) Except when all MFIVs, MFRVs, and associated bypass valves are closed and de-activated or isolated by a closed manual valve. (g) MODE 2 if Turbine Driven Main Feed Pumps are operating.  
ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 3 of 7)
(h) For the time period between February 23, 2000, and prior to turbine restart (following the next time the turbine is removed from service), the response time test requirement of SR 3.3.2.10 is not applicable for 1-FSV-47-027.
Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES                                                                             NOMINAL OR OTHER SPECIFIED             REQUIRED                  SURVEILLANCE      ALLOWABLE            TRIP FUNCTION              CONDITIONS                 CHANNELS       CONDITIONS REQUIREMENTS         VALUE         SETPOINT
: 4. Steam Line Isolation (continued)
(c)    (c)
: c. Containment           1, 2 , 3                       4             E       SR 3.3.2.1         2.9 psig       2.8 psig Pressure-                                                                   SR 3.3.2.4 High High                                                                   SR 3.3.2.9 SR 3.3.2.10
: d. Steam Line Pressure (c)
(1)   Low             1, 2     , 3(a) (c)       3 per steam         D       SR 3.3.2.1       666.6 (b) psig 675 (b) psig line                   SR 3.3.2.4 SR 3.3.2.9 SR .3.2.10 (d) (c)                                                                  (e)           (e)
(2)   Negative            3                      3 per steam         D       SR 3.3.2.1         108.5     psi   100     psi Rate-High                                     line                   SR 3.3.2.4 SR 3.3.2.9 SR 3.3.2.10
: 5. Turbine Trip and Feedwater Isolation (f)
: a. Automatic               1, 2     , 3(f)           2 trains         H       SR 3.3.2.2           NA               NA Actuation Logic                                                             SR 3.3.2.3 and Actuation                                                               SR 3.3.2.5 Relays (f)
: b. SG Water               1, 2     , 3(f)           3 per SG           I       SR 3.3.2.1           83.1%           82.4%
Level-High                                                                   SR 3.3.2.4 High(P-14)                                                                   SR 3.3.2.9 (h)
SR 3.3.2.10
: c. Safety           Refer to Function 1 (Safety Injection) for all initiation Injection       functions and requirements.
(f) (g)
: d. North MSV Vault                 1, 2 ,                 3/vault         O     SR 3.3.2.6         5.31 inches       4 inches Room Water                                               Room                   SR 3.3.2.9 Level - High (f), (g)
: e. South MSV Vault                 1, 2                   3/vault         O     SR 3.3.2.6         4.56 inches       4 inches Room Water                                               Room                   SR 3.3.2.9 Level - High (continued)
(a) Above the P-11 (Pressurizer Pressure) interlock.
(b) Time constants used in the lead/lag controller are t1 50 seconds and t2 5 seconds.
(c) Except when all MSIVs are closed and de-activated.
(d) Function automatically blocked above P-11 (Pressurizer Interlock) setpoint and is enabled below P-11 when safety injection on Steam Line Pressure Low is manually blocked.
(e) Time constants utilized in the rate/lag controller are t3 and t4 50 seconds.
(f) Except when all MFIVs, MFRVs, and associated bypass valves are closed and de-activated or isolated by a closed manual valve.
(g) MODE 2 if Turbine Driven Main Feed Pumps are operating.
(h) For the time period between February 23, 2000, and prior to turbine restart (following the next time the turbine is removed from service), the response time test requirement of SR 3.3.2.10 is not applicable for 1-FSV-47-027.
Watts Bar-Unit 1                                                    3.3-36                                      Amendment 23


ESFAS Instrumentation B 3.3.2 BASES     (continued)    Watts Bar-Unit 1 B 3.3-104 Revision 90  Amendment 68    ACTIONS D.1, D.2.1, and D.2.2 (continued)     Failure to restore the inoperable channel to OPERABLE status or place it in the tripped condition within 72 hours requires the plant be placed in MODE 3 within the following 6 hours and MODE 4 within the next 6 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODE 4, these Functions are no longer required OPERABLE. The Required Actions have been modified by a Note that allows placing an inoperable channel in the bypassed condition for up to 12 hours while performing routine surveillance testing of other channels. The Note also allows a channel to be placed in bypass for up to 12 hours for testing of the bypassed channel. However, only one channel may be placed in bypass at any one time. The 12 hours allowed for testing are justified in Reference 17.     E.1, E.2.1, and E.2.2     Condition E applies to:     Containment Spray Containment Pressure-High High;       Steam Line Isolation Containment Pressure-High High;     and     Containment Phase B Isolation Containment Pressure-     High High. None of these signals has input to a control function. Thus, two-out-of-three logic is necessary to meet acceptable protective requirements. However, a two-out-of-three design would require tripping a failed channel. This is undesirable because a single failure would then cause spurious containment spray initiation. Spurious spray actuation is undesirable because of the cleanup problems presented. Therefore, these channels are designed with ESFAS Instrumentation B 3.3.2  BASES    (continued)   Watts Bar-Unit 1 B 3.3-105 Revision 90   Amendment 68     ACTIONS E.1, E.2.1, and E.2.2 (continued)       two-out-of-four logic so that a failed channel may be bypassed rather than tripped. Note that one channel may be bypassed and still satisfy the single failure criterion. Furthermore, with one channel bypassed, a single instrumentation channel failure will not spuriously initiate containment spray. To avoid the inadvertent actuation of containment spray and Phase B containment isolation, the inoperable channel should not be placed in the tripped condition. Instead it is bypassed. Restoring the channel to OPERABLE status, or placing the inoperable channel in the bypass condition within 72 hours, is sufficient to assure that the Function remains OPERABLE and minimizes the time that the Function may be in a partial trip condition (assuming the inoperable channel has failed high). The Completion Time is further justified based on the low probability of an event occurring during this interval. Failure to restore the inoperable channel to OPERABLE status, or place it in the bypassed condition within 72 hours, requires the plant be placed in MODE 3 within the following 6 hours and MODE 4 within the next 6 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODE 4, these Functions are no longer required OPERABLE. The Required Actions are modified by a Note that allows placing one channel in bypass for up to 12 hours while performing routine surveillance testing. The channel to be tested can be tested in bypass with the inoperable channel also in bypass. The time limit is justified in Reference 17.     F.1, F.2.1, and F.2.2     Condition F applies to:     Manual Initiation of Steam Line Isolation;     Loss of Offsite Power;     Auxiliary Feedwater Pump Suction Transfer on Suction PressureLow; and Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 3 of 16  II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]: A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)] Some examples of SRO exam items for this topic include:
ESFAS Instrumentation B 3.3.2 BASES ACTIONS         D.1, D.2.1, and D.2.2 (continued)
Failure to restore the inoperable channel to OPERABLE status or place it in the tripped condition within 72 hours requires the plant be placed in MODE 3 within the following 6 hours and MODE 4 within the next 6 hours.
The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODE 4, these Functions are no longer required OPERABLE.
The Required Actions have been modified by a Note that allows placing an inoperable channel in the bypassed condition for up to 12 hours while performing routine surveillance testing of other channels. The Note also allows a channel to be placed in bypass for up to 12 hours for testing of the bypassed channel.
However, only one channel may be placed in bypass at any one time. The 12 hours allowed for testing are justified in Reference 17.
E.1, E.2.1, and E.2.2 Condition E applies to:
Containment Spray Containment Pressure-High High; Steam Line Isolation Containment Pressure-High High; and Containment Phase B Isolation Containment Pressure-High High.
None of these signals has input to a control function. Thus, two-out-of-three logic is necessary to meet acceptable protective requirements. However, a two-out-of-three design would require tripping a failed channel. This is undesirable because a single failure would then cause spurious containment spray initiation. Spurious spray actuation is undesirable because of the cleanup problems presented. Therefore, these channels are designed with (continued)
Watts Bar-Unit 1                         B 3.3-104                                      Revision 90 Amendment 68
 
ESFAS Instrumentation B 3.3.2 BASES ACTIONS         E.1, E.2.1, and E.2.2 (continued) two-out-of-four logic so that a failed channel may be bypassed rather than tripped. Note that one channel may be bypassed and still satisfy the single failure criterion. Furthermore, with one channel bypassed, a single instrumentation channel failure will not spuriously initiate containment spray.
To avoid the inadvertent actuation of containment spray and Phase B containment isolation, the inoperable channel should not be placed in the tripped condition. Instead it is bypassed. Restoring the channel to OPERABLE status, or placing the inoperable channel in the bypass condition within 72 hours, is sufficient to assure that the Function remains OPERABLE and minimizes the time that the Function may be in a partial trip condition (assuming the inoperable channel has failed high). The Completion Time is further justified based on the low probability of an event occurring during this interval. Failure to restore the inoperable channel to OPERABLE status, or place it in the bypassed condition within 72 hours, requires the plant be placed in MODE 3 within the following 6 hours and MODE 4 within the next 6 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODE 4, these Functions are no longer required OPERABLE.
The Required Actions are modified by a Note that allows placing one channel in bypass for up to 12 hours while performing routine surveillance testing. The channel to be tested can be tested in bypass with the inoperable channel also in bypass. The time limit is justified in Reference 17.
F.1, F.2.1, and F.2.2 Condition F applies to:
Manual Initiation of Steam Line Isolation; Loss of Offsite Power; Auxiliary Feedwater Pump Suction Transfer on Suction PressureLow; and (continued)
Watts Bar-Unit 1                          B 3.3-105                                      Revision 90 Amendment 68
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]:
A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)]
Some examples of SRO exam items for this topic include:
* Reporting requirements when the maximum licensed thermal power output is exceeded.
* Reporting requirements when the maximum licensed thermal power output is exceeded.
* Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems, fire doors, etc.
* Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems, fire doors, etc.
* The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
* The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
* National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
* National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
* Processes for TS and FSAR changes. Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic. B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)] Some examples of SRO exam items for this topic include:
* Processes for TS and FSAR changes.
Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic.
B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]
Some examples of SRO exam items for this topic include:
* Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
* Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
* Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
* Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
* Knowledge of TS bases that are required to analyze TS required actions and terminology.
* Knowledge of TS bases that are required to analyze TS required actions and terminology.
* Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).
* Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of  1 hour action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.  
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of  1 hour action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. ROs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Page 3 of 16


SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. RO's are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 4 of 16 ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)         Above this line RO knowledge Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 5 of 16 Figure 1:  Screening for SRO-only linked to 10 CFR 55.43(b)(2) (Tech Specs)       
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)
The knowledge required to complete this question is "below the line."
RO knowledge Above this line Page 4 of 16


Can question be answered solely by knowing  1 hour TS/TRM Action? RO question YesNoCan question be answered solely by knowing the LCO/TRM information listed "above-the-line?" YesRO question NoCan question be answered solely by knowing the TS Safety Limits? YesRO question NoDoes the question involve one or more of the following for TS, TRM, or ODCM?
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)
(Tech Specs)
Can question be answered solely by knowing  1       Yes hour TS/TRM Action?                                           RO question No Can question be answered solely by knowing the       Yes LCO/TRM information listed above-the-line?                   RO question No Can question be answered solely by knowing the       Yes TS Safety Limits?                                             RO question No Does the question involve one or more of the following for TS, TRM, or ODCM?
* Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
* Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)
* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)                             Yes      SRO-only
* Knowledge of TS bases that is required to analyze TS required actions and terminology SRO-only question YesNoQuestion might not be linked to 10 CFR 55.43(b)(2) for SRO-only
* Knowledge of TS bases that is required to analyze TS                   question required actions and terminology No Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only Page 5 of 16
: 91. Given the following timeline: 00:00:00 - Unit 1 is at 100% power. 00:0 1 :00 FR-S.1, Nuclear Power Generation/ATWS is entered. 00: 1 5:00 - In accordance with step 19, the crew CHECKS Incore T/Cs. - ALL T/Cs are greater than (>) 1200&deg;F slowly RISING  - ALL Power Range Detectors indicate 4%. - ALL Intermediate Range SUR Monitors indicate -0.1 dpm. Which ONE of the following completes the statements below?  At 00: 1 5:00, T (T = twall - tcc), as shown in the picture above, is ____(1)____ it was at 00:00:00. At 00: 1 5:0 1, in accordance with 1-FR-S.1, the crew will be required to GO TO _____(2)_____. A. (1) the same as (2) SACRG-1, Severe Accident Control Room Guideline Initial Response B. (1) greater than (2) SACRG-1, Severe Accident Control Room Guideline Initial Response C. (1) the same as (2) 1-FR-C.1, Inadequate Core Cooling D. (1) greater than (2) 1-FR-C.1, Inadequate Core Cooling CORRECT ANSWER:B DISTRACTOR ANALYSIS:  A. Incorrect: WCAP-9753, "Inadequate Core Cooling Studies of Scenarios With Feedwater Available, Using the NOTRUMP Computer Code," was produced by Westinghouse and demonstrated that while the bulk fluid temperature (as measured by the incore thermocouples) was not equal to the fuel metal temperature it was related to such.
As seen in the Westinghouse background document for the status tree F-0.2, "Core Cooling."  the 1200&deg;F setpoint for incore temperature "indicates that most liquid inventory has already been removed from the RCS and that core decay heat is superheating steam in the core."  The study, "Thermal-hydraulic behavior of electrically heated [rod] during a critical heat flux transient," demonstrates that once the heat flux present within a process flowpath increases such that nucleate boiling can no longer exist, the heat transfer coefficient between the electrically heated rod (e.g. nuclear fuel rod) and the process flow (e.g. Reactor Coolant System fluid) drops dramatically. Also demonstrated (and thus expected) is the fact that the surface temperature of the electrically heated rod rises incredibly (grossly in excess of that of the bulk-albeit superheated-coolant). 


WCAP-9753 further demonstrated for given cases of depleted coolant inventories that when the incore T/Cs read 1200&deg;F, the fuel metal temperature reached temperatures over 2000 &deg;F. This indicated a T (between the wall of the clad and the bulk coolant) of over 800&deg;F. Depending upon which fuel location was chosen-the temperature difference could be much higher.  
91.
Given the following timeline:
00:00:00 -        Unit 1 is at 100% power.
00:0 1 :00 -      1-FR-S.1, Nuclear Power Generation/ATWS is entered.
00: 1 5:00 -     In accordance with step 19, the crew CHECKS Incore T/Cs.
                  -  ALL T/Cs are greater than (>) 1200&deg;F slowly RISING .
                  -  ALL Power Range Detectors indicate 4%.
                  -  ALL Intermediate Range SUR Monitors indicate -0.1 dpm.
Which ONE of the following completes the statements below?
At 00: 1 5:00, T (T = twall - tcc), as shown in the picture above, is ____(1)____ it was at 00:00:00.
At 00: 1 5:0 1, in accordance with 1-FR-S.1, the crew will be required to GO TO _____(2)_____.
A.      (1)      the same as (2)     SACRG-1, Severe Accident Control Room Guideline Initial Response B.     (1)      greater than (2)      SACRG-1, Severe Accident Control Room Guideline Initial Response C.      (1)      the same as (2)      1-FR-C.1, Inadequate Core Cooling D.      (1)      greater than (2)      1-FR-C.1, Inadequate Core Cooling


By contrast, at 100% power, section 4.4.2.2.5 of the UFSAR states that, "the outer surface of the fuel rod at the hot spot operates at a temperature of approximately 660&deg;F for steady state operation at rated power throughout core life due to the onset of nucleate boiling."  As can be found on the integrated computer system or on the RVLIS thermocouple displays themselves, the average of all of the incore thermocouples is approximately 620&deg;F. Therefore, the nominal difference between the hottest fuel rod's metal temperature and that registered by the incore thermocouples is approximately 40&deg;F at 100% power.
CORRECT ANSWER:                                                                B DISTRACTOR ANALYSIS:
Therefore, it is incorrect to believe that the T specified in the stem of the question remains the same between the two times. It is plausible to believe this given a failure to understand the heat transfer properties in effect for the fuel during this accident.  
A. Incorrect: WCAP-9753, Inadequate Core Cooling Studies of Scenarios With Feedwater Available, Using the NOTRUMP Computer Code, was produced by Westinghouse and demonstrated that while the bulk fluid temperature (as measured by the incore thermocouples) was not equal to the fuel metal temperature it was related to such.
As seen in the Westinghouse background document for the status tree F-0.2, Core Cooling. the 1200&deg;F setpoint for incore temperature indicates that most liquid inventory has already been removed from the RCS and that core decay heat is superheating steam in the core.
The study, Thermal-hydraulic behavior of electrically heated [rod] during a critical heat flux transient, demonstrates that once the heat flux present within a process flowpath increases such that nucleate boiling can no longer exist, the heat transfer coefficient between the electrically heated rod (e.g. nuclear fuel rod) and the process flow (e.g. Reactor Coolant System fluid) drops dramatically. Also demonstrated (and thus expected) is the fact that the surface temperature of the electrically heated rod rises incredibly (grossly in excess of that of the bulk-albeit superheated-coolant).
WCAP-9753 further demonstrated for given cases of depleted coolant inventories that when the incore T/Cs read 1200&deg;F, the fuel metal temperature reached temperatures over 2000 &deg;F. This indicated a T (between the wall of the clad and the bulk coolant) of over 800&deg;F. Depending upon which fuel location was chosen-the temperature difference could be much higher.
By contrast, at 100% power, section 4.4.2.2.5 of the UFSAR states that, the outer surface of the fuel rod at the hot spot operates at a temperature of approximately 660&deg;F for steady state operation at rated power throughout core life due to the onset of nucleate boiling. As can be found on the integrated computer system or on the RVLIS thermocouple displays themselves, the average of all of the incore thermocouples is approximately 620&deg;F. Therefore, the nominal difference between the hottest fuel rods metal temperature and that registered by the incore thermocouples is approximately 40&deg;F at 100% power.
Therefore, it is incorrect to believe that the T specified in the stem of the question remains the same between the two times. It is plausible to believe this given a failure to understand the heat transfer properties in effect for the fuel during this accident.
Step 19 of 1-FR-S.1 is as follows:
: 19. CHECK Incore T/Cs less than 1200&deg;F.
The response not obtained for step 19 is:
IF Incore T/Cs are greater than 1200&deg;F AND rising, THEN


Step 19 of 1-FR-S.1 is as follows: 19. CHECK Incore T/Cs less than 1200F. The response not obtained for step 19 is: IF Incore T/Cs are greater than 1200F AND rising, THEN 
              ** GO TO 1-SACRG-1, Severe Accident Control Room Guideline Initial Response.
** GO TO 1-SACRG-1, Severe Accident Control Room Guideline Initial Response.
Step 20 is:
Step 20 is:  
: 20. CHECK reactor subcritical:
: 20. CHECK reactor subcritical: a. Power range channels less than 5%.  
: a. Power range channels less than 5%.
: b. Intermediate range startup rate NEGATIVE.
: b. Intermediate range startup rate NEGATIVE.
Contained within the step 20 response not obtained is: IF red OR orange condition exists on other Status Trees, THEN PERFORM actions of other FR Procedures which do not cool down or otherwise add positive reactivity to the core.  
Contained within the step 20 response not obtained is:
IF red OR orange condition exists on other Status Trees, THEN PERFORM actions of other FR Procedures which do not cool down or otherwise add positive reactivity to the core.
Therefore, it is correct that a transition to 1-SACRG-1 is warranted.
B. Correct: It is correct that at 00:15:00, the T specified in the stem of the question is greater than it was at 00:00:00. Also, it is correct to transition to 1-SACRG-1.
C. Incorrect: Again, it is incorrect that the T specified in the stem of the question remains the same. Also, it is incorrect that a transition to 1-FR-C.1 would be required. It is plausible to believe this as if one misses the transition to 1-SACRG-1; one will arrive at step 20 with the conditions met for a RED path to 1-FR-C.1. One could utilize the response not obtained for step 20 to perform actions of other FR procedures (in this case 1-FR-C.1).
D. Incorrect: While it is correct that at 00:15:00, the T specified in the stem of the question is greater than it was at 00:00:00. It is not correct that a transition to 1-FR-C.1 would be required.
 
Question Number:        91 Tier:    2  Group:        2 K/A:    017 In-Core Temperature Monitor (ITM) System A2 Ability to (a) predict the impacts of the following malfunctions or operations on the ITM system; and (b) based on those predictions, use procedures to Correct: control or mitigate the consequences of those malfunctions or operations:
A2.02 Core damage Importance Rating:      3.6 4.1 10 CFR Part 55:        (CFR: 41.5 / 43.5 / 45.3 / 45.5) 10CFR55.43.b:        10 CFR 55.43(b)(5)
K/A Match:    The K/A is matched because given a case of core damage (the conditions met for entry into 1-SACRG-1), the applicant must understand the relation of the temperature indicated by the ITM to that of the fuel metal. Next, the applicant must correctly implement the functional restoration procedures to correctly transition to 1-SACRG-1 in order that the core damage (and high temperatures of ITM) be mitigated.
Technical


Therefore, it is correct that a transition to 1-SACRG-1 is warranted. B. Correct: It is correct that at 00:15:00, the T specified in the stem of the question is greater than it was at 00:00:00. Also, it is correct to transition to 1-SACRG-1. C. Incorrect: Again, it is incorrect that the T specified in the stem of the question remains the same. Also, it is incorrect that a transition to 1-FR-C.1 would be required. It is plausible to believe this as if one misses the transition to 1-SACRG-1; one will arrive at step 20 with the conditions met for a RED path to 1-FR-C.1. One could utilize the response not obtained for step 20 to perform actions of other FR procedures (in this case 1-FR-C.1). D. Incorrect: While it is correct that at 00:15:00, the T specified in the stem of the question is greater than it was at 00:00:00. It is not correct that a transition to 1-FR-C.1 would be required.
==Reference:==
Question Number: 91  Tier:   2 Group:   2 K/A: 017 In-Core Temperature Monitor (ITM) System A2 Ability to (a) predict the impacts of the following malfunctions or operations on the ITM system; and (b) based on those predictions, use procedures to Correct: control or mitigate the consequences of those malfunctions or operations:
Section 4.4.2.2.5 of the UFSAR ICS screen shot showing the incore T/C temperatures at 100% power WCAP-9753, Inadequate Core Cooling Studies of Scenarios With Feedwater Available, Using the NOTRUMP Computer Code.
A2.02 Core damage Importance Rating: 3.6  4.1  10 CFR Part 55: (CFR: 41.5 / 43.5 / 45.3 / 45.5)
Thermal-hydraulic behavior of electrically heated [rod] during a critical heat flux transient" Westinghouse owners group background document for status tree F-0.2, Core Cooling" 1-FR-S.1, Nuclear Power Generation/ATWS" Proposed references to     None be provided:
Learning Objective:        3-OT-FRS0001
: 9. Given a set of plant conditions, diagnose and implement action steps, RNOs, notes and cautions of 1-FR-S.1 Cognitive Level:
Higher              X Lower Question Source:
New                  X Modified Bank Bank Question History:           New question for the 2015-301 NRC SRO Exam Comments:                   See the attached marked up copy of the SRO ONLY guidance.


10CFR55.43.b: 10 CFR 55.43(b)(5)  
WESTINGHOUSE PROPRIETARY CLASS 2 feedwater. The two effects combine to sharply depressurize the secon-dary (Figures 86 and 87) until, at 600 psia secondary pressure, the steam dump system is automatically isolated. These events occurring on the secondary side do not noticeably alter conditions in the primary loops, since the heat transfer coefficient between the steam generator tubes and the superheated steam on the primary side is relatively small, and the secondary to primary heat transfer is low.
At 1700 seconds when the maximum average fuel temperature of about 2200&deg;F is attained (Figure 107), primary pressure drops below the low head safety injection shut-off head, initiating core recovery. Although the core recovers in this case with minimal operator actions, higher peak temperatures would be expected than are indicated by this analysis, since only the average core assembly is modeled and zirc-water reactions are neg 1ected.
: 3. Core Exit Thermocouple Response In the one inch break transient, considerable recirculation was found to occur between the upper core node, the upper plenum and the guide
~ubes. Three dimensional considerations, however, lead to the conclu-sion that the core exit thermocouples would indicate the core exit tem-perature under inadequate core cooling conditions.
For the four inch break case, however, since all core flows are upward and out the top of the core, the upper core node temperature would be expected to give a good estimate of the thermocouple reading. Note that at 1363 seconds, the upper core node temperature reaches 1200&deg;F, before the core has completely uncovered.
: 4. Hot Leg Temperature Response Figures 102 and 103 present the response of the resistance thermometers in the broken and intact loop hot legs, respectively. While the broken loop hot leg temperature shows only a slight rise at about the time the 17


K/A Match: The K/A is matched because given a case of core damage (the conditions met for entry into 1-SACRG-1), the applicant must understand the relation of the temperature indicated by the ITM to that of the fuel metal. Next, the applicant must correctly implement the functional restoration procedures to correctly transition to 1-SACRG-1 in order that the core damage (and high temperatures of ITM) be mitigated. Technical
BLOCK DESCRIPTION TABLE FOR STATUS TREE F-0.2 BLOCK DECISION:    Core Exit TCs Less Than 1200&deg;F PURPOSE:  To determine if inadequate core cooling has been reached BASIS:
Analyses of inadequate core cooling scenarios (References 1 and 2) show that core exit temperature greater than 1200&deg;F is a satisfactory criterion for basing extreme operator action. At least 5 thermocouples should be reading greater than 1200&deg;F. Five has been chosen to allow for thermocouples failing high. This temperature indicates that most liquid inventory has already been removed from the RCS and that core decay heat is superheating steam in the core. An extreme challenge to the fuel matrix/clad barrier is imminent and a RED priority is warranted. The appropriate guideline for functional response is FR-C.1, RESPONSE TO INADEQUATE CORE COOLING. If core exit thermocouples are less than 1200&deg;F, then subsequent blocks check for other extreme, severe, not satisfied or satisfied conditions for the safety function.
INSTRUMENTATION:
Decay heat removal still exists Core exit thermocouples temperature indication      Core damage will (even if it is a very result if this poor form of such).
KNOWLEDGE:                                          "extreme challenge" is not resolved (i.e. via N/A implementation of 1-FR-C.1).
PLANT-SPECIFIC INFORMATION:
o    The following criteria should be used to determine which thermocouples to monitor:
: 1) At least one thermocouple should be located as close as possible to the geometric center of the core.
: 2) The other thermocouples should be located at least one per quadrant over the highest power assemblies in each quadrant. The outer two rows of assemblies should be excluded, since they can receive significant cooling from steam generator drainage due to refluxing. The thermocouples should be selected at each refueling to ensure that the highest power assemblies are always being used.
F-0.2 Background                    6              HP/LP-Rev. 2, 4/30/2005 HF02BG.doc


==Reference:==
Thermalhydraulic Behavior of Electrically Heated Rod during a Critical Heat Flux Transient Rita de C&#xe1;ssia Fernandes de Lima Mechanical Engineering Department. Universidade Federal de Pernambuco Av. Acad. H&#xe9;lio Ramos, s/n 50740-530 Recife. PE. Brazil ritalima@npd.ufpe.br Pedro Carajilescov Mechanical Engineering Department. Universidade Federal Fluminense Rua Passo da P&#xe1;tria,156 24210-240 Niter&#xf3;i. RJ. Brazil pedroc@caa.uff.br In nuclear reactors, the occurrence of critical heat flux leads to fuel rod overheating with clad fusion and radioactive products leakage. To predict the effects of such phenomenon, experiments are performed using electrically heated rods to simulate operational and accidental conditions of nuclear fuel rods. In the present work, it is performed a theoretical analysis of the drying and rewetting front propagation during a critical heat flux experiment, starting with the application of an electrical power step from steady state condition. After the occurrence of critical heat flux, the drying front propagation is predicted. After a few seconds, a power cut is considered and the rewetting front behavior is analytically observed. Studies performed with various values of coolant mass flow rate show that this variable has more influence on the drying front velocity than on the rewetting one.
Section 4.4.2.2.5 of the UFSAR ICS screen shot showing the incore T/C temperatures at 100% power WCAP-9753, "Inadequate Core Cooling Studies of Scenarios With Feedwater Available, Using the NOTRUMP Computer Code." "Thermal-hydraulic behavior of electrically heated [rod] during a critical heat flux transient" Westinghouse owners' group background document for status tree F-0.2, "Core Cooling" 1-FR-S.1, "Nuclear Power Generation/ATWS"  Proposed references to be provided: None  Learning Objective: 3-OT-FRS0001 9. Given a set of plant conditions, diagnose and implement action steps, RNOs, notes and cautions of 1-FR-S.1  Cognitive Level:      Higher X  Lower    Question Source:      New X  Modified Bank  Bank      Question History: New question for the 2015-301 NRC SRO Exam  Comments: See the attached marked up copy of the SRO ONLY guidance.
Keywords: Critical heat flux, rewetting front, drying front, thermalhydraulics, numerical simulation Introduction The power generation of a nuclear reactor is limited by the coolant capability of removing the heat generated inside the fuel rods. In PWR type reactor, this capability is determined by the occurrence of critical heat flux (CHF),
WESTINGHOUSE PROPRIETARY CLASS 2 feedwater. The two effects combine to sharply depressurize the dary (Figures 86 and 87) until, at 600 psia secondary pressure, the steam dump system is automatically isolated. These events occurring on the secondary side do not noticeably alter conditions in the primary loops, since the heat transfer coefficient between the steam generator tubes and the superheated steam on the primary side is relatively small, and the secondary to primary heat transfer is low. At 1700 seconds when the maximum average fuel temperature of about 2200&deg;F is attained (Figure 107), primary pressure drops below the low head safety injection shut-off head, initiating core recovery. Although the core recovers in this case with minimal operator actions, higher peak temperatures would be expected than are indicated by this analysis, since only the average core assembly is modeled and zirc-water reactions are neg 1 ected. 3. Core Exit Thermocouple Response In the one inch break transient, considerable recirculation was found to occur between the upper core node, the upper plenum and the guide Three dimensional considerations, however, lead to the sion that the core exit thermocouples would indicate the core exit perature under inadequate core cooling conditions. For the four inch break case, however, since all core flows are upward and out the top of the core, the upper core node temperature would be expected to give a good estimate of the thermocouple reading. Note that at 1363 seconds, the upper core node temperature reaches 1200&deg;F, before the core has completely uncovered. 4. Hot Leg Temperature Response Figures 102 and 103 present the response of the resistance thermometers in the broken and intact loop hot legs, respectively. While the broken loop hot leg temperature shows only a slight rise at about the time the 17 BLOCKDECISION:BLOCKDESCRIPTIONTABLEFORSTATUSTREEF-0.2CoreExitTCsLessThan1200&deg;FPURPOSE:TodetermineifinadequatecorecoolinghasbeenreachedBASIS:Analysesofinadequatecorecoolingscenarios(References1and2)showthatcoreexittemperaturegreaterthan1200&deg;Fisasatisfactorycriterionforbasingextremeoperatoraction.Atleast5thermocouplesshouldbereadinggreaterthan1200&deg;F.Fivehasbeenchosentoallowforthermocouplesfailinghigh.ThistemperatureindicatesthatmostliquidinventoryhasalreadybeenremovedfromtheRCSandthatcoredecayheatissuperheatingsteaminthecore.Anextremechallengetothefuelmatrix/cladbarrierisimminentandaREDpriorityiswarranted.TheappropriateguidelineforfunctionalresponseisFR-C.1,RESPONSETOINADEQUATECORECOOLING.Ifcoreexitthermocouplesarelessthan1200&deg;F,thensubsequentblockscheckforotherextreme,severe,notsatisfiedorsatisfiedconditionsforthesafetyfunction.INSTRUMENTATION:CoreexitthermocouplestemperatureindicationKNOWLEDGE:N/APLANT-SPECIFICINFORMATION:oThefollowingcriteriashouldbeusedtodeterminewhichthermocouplestomonitor:1)Atleastonethermocoupleshouldbelocatedascloseaspossibletothegeometriccenterofthecore.2)Theotherthermocouplesshouldbelocatedatleastoneperquadrantoverthehighestpowerassembliesineachquadrant.Theoutertworowsofassembliesshouldbeexcluded,sincetheycanreceivesignificantcoolingfromsteamgeneratordrainageduetorefluxing.Thethermocouplesshouldbeselectedateachrefuelingtoensurethatthehighestpowerassembliesarealwaysbeingused.F-0.2BackgroundHF02BG.doc6HP/LP-Rev.2,4/30/2005 Thermalhydraulic Behavior of Electrically Heated Rod during a Critical Heat Flux Transient Rita de C&#xe1;ssia Fernandes de Lima  Mechanical Engineering Department. Universidade Federal de Pernambuco  Av. Acad. H&#xe9;lio Ramos, s/n 50740-530 Recife. PE. Brazil ritalima@npd.ufpe.br Pedro Carajilescov Mechanical Engineering Department. Universidade Federal Fluminense  Rua Passo da P&#xe1;tria,156 24210-240 Niter&#xf3;i. RJ. Brazil pedroc@caa.uff.br In nuclear reactors, the occurrence of critical heat flux leads to fuel rod overheating with clad fusion and radioactive products leakage. To predict the effects of such phenomenon, experiments are performed using electrically heated rods to simulate operational and accidental conditions of nuclear fuel rods. In the present work, it is performed a theoretical analysis of the drying and rewetting front propagation during a critical heat flux experiment, starting with the application of an electrical power step from steady state condition. After the occurrence of critical heat flux, the drying front propagation is predicted. After a few seconds, a power cut is considered and the rewetting front behavior is analytically observed. Studies performed with various values of coolant mass flow rate show that this variable has more influence on the drying front velocity than on the rewetting one. Keywords: Critical heat flux, rewetting front, drying front, thermalhydraulics, numerical simulation Introduction The power generation of a nuclear reactor is limited by the coolant capability of removing the heat generated inside the fuel rods. In PWR type reactor, this capability is determined by the occurrence of critical heat flux (CHF), also called by DNB (Departure from Nucleated Boiling). Fuel rods overheating, due to the occurrence of a boiling crisis, during a power transient, can yield clad fusion with radioactive products leakage to the coolant. To predict this type of phenomenon, it is a common practice to perform simulations of the reactor operational and transient conditions in thermalhydraulic loops, utilizing electrically heated rods. Such simulations represent an important aspect of the reactor safety analysis.
also called by DNB (Departure from Nucleated Boiling). Fuel rods overheating, due to the occurrence of a boiling crisis, during a power transient, can yield clad fusion with radioactive products leakage to the coolant. To predict this type of phenomenon, it is a common practice to perform simulations of the reactor operational and transient conditions in thermalhydraulic loops, utilizing electrically heated rods. Such simulations represent an important aspect of the reactor safety analysis.
When a critical heat flux occurs, the heated surface is covered by a vapor blanket, which will spread over the rod length with a so called drying front propagation velocity. Considering that vapor has a very low thermal conductivity, the local heat transfer coefficient is drastically reduced, provoking very high local temperatures. When the power is turned off, the wetting of the surface is re-established with a rewetting front propagation velocity. Since those electrically heated rods are very expensive, during a DNB experiment, it is necessary to turn the power off very quickly in order not to damage the rods. The prediction of the drying and rewetting front propagation, for a given experiment, can be used to establish the amount of time available for the experiment without rod damage. Several authors have studied this problem. Gunnerson & Yackle (1981) establish the difference between quench and rewetting. Yu et alii (1977) analyze the quench process on hot surfaces using bidimensional conduction. The work treats the subcooled and saturated rewetting for pressures from 1 to 69 bar. Olek et alii (1988) study the rewetting in descendent films. They consider the problem as a heat conjugated problem. Carlson (1989) analyses the expansion of the CHF region in direct heated rods. He takes into consideration also the thermocouple locations and the mechanism through which the drying front is detected by them. The present work, analyzes the thermal behavior of a typical electrically heated rod with indirect heating, as shown in Figure 1, during a step transient of the electrical power. From a steady state condition, at a given power, it is imposed a 10% power step of its initial value in order to produce the CHF. It is, then, observed the effect of pressure and flow rate in the drying and rewetting front propagation velocities.
 
Theoretical Model Consider the universally adopted test section, with indirect heating, shown schematically in Figure 1. Heat is generated in the electrical resistance by an electrical current, is conducted axially and radially, and is removed by the water flowing longitudinally along the rod. The heat transfer coefficient is a function of the local coolant conditions. In order to analyze the thermal behavior of this system, it was considered:  circular symmetry of the electrical heaters;  uniform axial heat generation;  no heat losses through the ends of the rod;  constant thermal properties of the materials of the rod;  homogeneous model for the water two-phase flow. For the several regions of the rod, the heat conduction equation can be written as:
When a critical heat flux occurs, the heated surface is covered by a vapor blanket, which will spread over the rod length with a so called drying front propagation velocity. Considering that vapor has a very low thermal conductivity, the local heat transfer coefficient is drastically reduced, provoking very high local temperatures. When the power is turned off, the wetting of the surface is re-established with a rewetting front propagation velocity. Since those electrically heated rods are very expensive, during a DNB experiment, it is necessary to turn the power off very quickly in order not to damage the rods. The prediction of the drying and rewetting front propagation, for a given experiment, can be used to establish the amount of time available for the experiment without rod damage. Several authors have studied this problem. Gunnerson & Yackle (1981) establish the difference between quench and rewetting. Yu et alii (1977) analyze the quench process on hot surfaces using bidimensional conduction. The work treats the subcooled and saturated rewetting for pressures from 1 to 69 bar. Olek et alii (1988) study the rewetting in descendent films. They consider the problem as a heat conjugated problem. Carlson (1989) analyses the expansion of the CHF region in direct heated rods. He takes into consideration also the thermocouple locations and the mechanism through which the drying front is detected by them.
where T is the rod temperature; k, the thermal conductivity;  , the density and cp , the specific heat of each material. The volumetric heat generation, q~~~, is zero for all material except for the electrical resistance. According to Silva Neto et alii(1983), the lumped form of the energy equation for the water coolant can be written as:  where q" is the heat flux received by the coolant; A is the cross section area of the channel; p is the rod perimeter and hf , f and G are the enthalpy, density and mass flux of the coolant, respectively. The coupling between the rod and the flow is established by the surface heat removal given by:  where kclad is the thermal conductivity of the cladding; href is the heat transfer coefficient between cladding and coolant; Tf is the coolant temperature and Rext =R4, the outer radius of the cladding. The correlations presented in Appendix A were considered for the heat transfer coefficient, taking into account the several heat transfer regimes. Although several different correlations can be found in literature, these were considered adequate for the present situation. The heat conduction equation for the rod was solved by the finite control volume method with an implicit formulation. For the water enthalpy, equations were solved iteratively. To take into account the surface-water coupling, in any length of the coolant channel, the sequence shown in Figure 2, with the variables Twall, href and q" defined in Appendix A, was adopted.
The present work, analyzes the thermal behavior of a typical electrically heated rod with indirect heating, as shown in Figure 1, during a step transient of the electrical power. From a steady state condition, at a given power, it is imposed a 10% power step of its initial value in order to produce the CHF. It is, then, observed the effect of pressure and flow rate in the drying and rewetting front propagation velocities.
Results A computational program was developed to analyze the critical heat flux for a several types of transients. Here the classical case of a step power transient is presented. At t=0s, the variable qo~ (the linear power density) is increased by 10% from its steady state value of 16 600 W/m, in order to reach the CHF. The new level is then maintained for 4.0 s and then the electric power is cut off. The entire transient lasts 4.5 s. The tables below show the physical and geometric parameters used. The electrically heated rod is composed of a Ni-Cr ( 18 to 20% Cr and 8 to 12% Ni ) resistance, a MgO electric insulator and stainless steel ( type 349 )
 
Theoretical Model Consider the universally adopted test section, with indirect heating, shown schematically in Figure 1. Heat is generated in the electrical resistance by an electrical current, is conducted axially and radially, and is removed by the water flowing longitudinally along the rod. The heat transfer coefficient is a function of the local coolant conditions.
In order to analyze the thermal behavior of this system, it was considered:
* circular symmetry of the electrical heaters;
* uniform axial heat generation;
* no heat losses through the ends of the rod;
* constant thermal properties of the materials of the rod;
* homogeneous model for the water two-phase flow.
For the several regions of the rod, the heat conduction equation can be written as:
 
where T is the rod temperature; k, the thermal conductivity;  , the density and cp , the specific heat of each material. The volumetric heat generation, q      , is zero for all material except for the electrical resistance.
According to Silva Neto et alii(1983), the lumped form of the energy equation for the water coolant can be written as:
where q" is the heat flux received by the coolant; A is the cross section area of the channel; p is the rod perimeter and hf , f and G are the enthalpy, density and mass flux of the coolant, respectively.
The coupling between the rod and the flow is established by the surface heat removal given by:
where kclad is the thermal conductivity of the cladding; href is the heat transfer coefficient between cladding and coolant; Tf is the coolant temperature and Rext =R4, the outer radius of the cladding.
The correlations presented in Appendix A were considered for the heat transfer coefficient, taking into account the several heat transfer regimes.
Although several different correlations can be found in literature, these were considered adequate for the present situation.
The heat conduction equation for the rod was solved by the finite control volume method with an implicit formulation. For the water enthalpy, equations were solved iteratively.
To take into account the surface-water coupling, in any length of the coolant channel, the sequence shown in Figure 2, with the variables Twall, href and q" defined in Appendix A, was adopted.
 
Results A computational program was developed to analyze the critical heat flux for a several types of transients. Here the classical case of a step power transient is presented. At t=0s, the variable qo (the linear power density) is increased by 10% from its steady state value of 16 600 W/m, in order to reach the CHF. The new level is then maintained for 4.0 s and then the electric power is cut off. The entire transient lasts 4.5 s.
The tables below show the physical and geometric parameters used. The electrically heated rod is composed of a Ni-Cr ( 18 to 20% Cr and 8 to 12%
Ni ) resistance, a MgO electric insulator and stainless steel ( type 349 )
cladding. The time step used in discretization is 5 x 10 -2 s. The numbers of nodes considered are: five in the internal insulator, nine in the resistance, four in the internal insulator and four in the cladding. The axial interval is equal to 10-2 m.
cladding. The time step used in discretization is 5 x 10 -2 s. The numbers of nodes considered are: five in the internal insulator, nine in the resistance, four in the internal insulator and four in the cladding. The axial interval is equal to 10-2 m.
The variation of the heat transfer coefficient versus height in the coolant channel, for several time instants, is shown in Fig. 3. Since the heat transfer coefficient for subcooled boiling is function of the wall temperature, it is observed that it will rise steadily until saturation is reached. After saturation, its value remains constant. When CHF occurs near the end of the channel, the heat removal degrades and the heat transfer coefficient suffers a severe drop as observed. This heat transfer crisis tends to travel to lower heights as time increases, which corresponds to a drying front propagation. For t = 4.0 s, the heat transfer coefficient drop is as large as 97%. When the power is cut off, the heat transfer coefficient drops along the rod due to the reduction of the rod superficial temperature.
The variation of the heat transfer coefficient versus height in the coolant channel, for several time instants, is shown in Fig. 3. Since the heat transfer coefficient for subcooled boiling is function of the wall temperature, it is observed that it will rise steadily until saturation is reached. After saturation, its value remains constant. When CHF occurs near the end of the channel, the heat removal degrades and the heat transfer coefficient suffers a severe drop as observed. This heat transfer crisis tends to travel to lower heights as time increases, which corresponds to a drying front propagation. For t = 4.0 s, the heat transfer coefficient drop is as large as 97%. When the power is cut off, the heat transfer coefficient drops along the rod due to the reduction of the rod superficial temperature.
Figure 4 shows the flow quality, where three different boiling regions can be observed, separated by the inflections of the curves. The first one separates the region of forced convection and subcooled boiling from the region of saturated boiling. The second inflection divides this last region from the post-dryout region. Near the entrance, a sharp increase in flow quality is observed, followed by a smooth increase when the CHF phenomenon occurs. There is a reduction in the quality growth, indicating the position and instant of time where it takes place. This behavior can be explained by the reduction of the superficial flux and consequently a smaller increase of the enthalpy. The maximum flow quality in this transient is equal to 0.57 at the outlet of the channel at t = 4.0 s.
 
Figure 4 shows the flow quality, where three different boiling regions can be observed, separated by the inflections of the curves. The first one separates the region of forced convection and subcooled boiling from the region of saturated boiling. The second inflection divides this last region from the post-dryout region. Near the entrance, a sharp increase in flow quality is observed, followed by a smooth increase when the CHF phenomenon occurs.
There is a reduction in the quality growth, indicating the position and instant of time where it takes place. This behavior can be explained by the reduction of the superficial flux and consequently a smaller increase of the enthalpy.
The maximum flow quality in this transient is equal to 0.57 at the outlet of the channel at t = 4.0 s.
 
The effect of the heat transfer regimes reflects on the clad temperature as shown in Fig.5. A maximum increase of 22% is seen in the clad temperatures when the boiling crisis phenomenon occurs. Clad melting can be avoided if the electric power supply is interrupted. Some clad points show temperature rise of 118 oC/s. As a result their temperatures could reach values as high as 1100oC in less than 7.0 s.
The effect of the heat transfer regimes reflects on the clad temperature as shown in Fig.5. A maximum increase of 22% is seen in the clad temperatures when the boiling crisis phenomenon occurs. Clad melting can be avoided if the electric power supply is interrupted. Some clad points show temperature rise of 118 oC/s. As a result their temperatures could reach values as high as 1100oC in less than 7.0 s.
The propagation of the drying and rewetting fronts is represented in Figure 6, for several mass flow rates. For mass flow rate equal to 0.0535 m/s, the drying front has a mean velocity of 4.6 cm/s. For the rewetting one, this velocity is 2.4 cm/s. The front velocities presented in Figs.6, 7 e 8 are mean velocities which are calculated dividing the maximum distance reached by the front by the correspondent time interval. The mass flow influence on the velocities is also shown. Variations of + 20% and - 20% on the reference case (= 0.0535 kg/s) are applied. It is noted that, as this variable rises, the drying velocity also rises from 4.6 to 4.9 cm/s, while the rewetting one goes from 2.4 to 4.4 cm/s. The CHF does not occur immediately after the power step, due to the radial and axial thermal resistances and condutances of the indirect heated rod. The time delay observed is reduced from 1.0 s to 0.8 s as the mass flow rate increases. 


In order to observe the pressure influence on the reported velocities, the same transient is then analyzed under the coolant pressure of 8.0 MPa. Its inlet temperature is now equal to 280 oC. A 41% reduction in the pressure value has a strong influence on the rewetting front velocity: an increase of 936%. Otherwise there is a little influence on the drying one: it goes from 3.8 to 3.2 cm/s. These comparisons are shown in Fig. 7. The pressure also has considerable influence on the time delay which varies from 1.9 s to 1.4 s as the pressure changes from 8.0 MPa to 13.5 MPa. The influence of the inlet mass flow rate was also investigated for the 8.0 MPa coolant pressure. Other authors have shown that this parameter has accentuated influence on the rewetting velocity at low pressures until 6.9 MPa. The test with a lower pressure was done to validate the model. The result obtained with the present model shown in Fig. 8 and confirms the trend. Note that the inlet mass flow rate has a smaller effect in the drying front. For a 20% reduction in mass flow the rewetting front velocity is reduced in 29.5%: It varies from 24.7 to 17.4 cm/s. Conclusions The present work analyzes the front propagation velocity for the drying out and rewetting processes, during the occurrence of critical heat flux in electrically heated simulators of nuclear fuel rods, caused by a power step. This study is very important in the simulation of nuclear power plants as well as in metallurgical problems. At the occurrence of CHF, the amount of time required to cut off the electric power used in the heating of the simulator needs to be quantified. After the p ower cut off, the surface is rewetted when the temperature of the wall is less than the critical one. The two phenomena were analyzed individually by several authors (Carlson (1989), Olek et al. (1988), Griffith et al.(1988)) and there were no information about the amount of time available for the operation of the protection systems. Specially, there were few informations about rewetting, studied before only for descending films at low pressures. The work here presented supplies part of this lack of information. Due to the radial and axial thermal resistances and capacitances of the indirect heating of the rod, the critical heat flux does not occur immediately after the power step. A certain time delay is observed. This time delay is reduced by increasing the pressure or the mass flow rate. In the beginning of the occurrence of CHF, the velocity of the propagation of the drying front is very high, being reduced gradually to an approximately constant value, around 4.6 cm/s. After the power cut off, the rewetting front presents a very large velocity of propagation, which is greatly affected by the system pressure and mass flow rate. The rewetting velocities were 24.7 cm/s and 2.4 cm/s for pressures of 8.0 MPa and 13.5 MPa, respectively. For the case of pressure of 8.0 MPa, the rewetting front propagation velocities were 17.4 cm/s and 24.7 cm/s for flow rates of 0.0471 kg/s and 0.0539 kg/s, respectively. At the spot where the occurrence of CHF first starts, it was observed that the temperature increases at a rate of 118&deg; C/s, which indicates that the wall temperature would reach its temperature limit, estimated around 1100&deg; C, in approximately 7s. This is the amount of time available to turn off the electrical power supply. This observed heating rate is much larger than the value obtain by Mosaad (1988). Additional details of the present work can be obtained in Lima (1997). Acknowledgements The authors thank to FACEPE ( Funda&#xe7;o de Apoio  Cincia e Tecnologia do Estado de Pernambuco) for the support given to this work. Appendix A a) Forced convection: DITTUS - BOELTER~s correlation (BJORNARD (1977)). where k is the coolant thermal conductivity; De , the hydraulic diameter of the channel; Re, the Reynolds number; and Pr , the Prandtl one. b) Nucleate boiling: THOM~s correlation (TONG & WEISMAN, (1979)). where Twall = external temperature of the cladding; Tsat = saturation temperature; q"SUP = heat flux; e p = pressure. The transition between forced convection and nucleate boiling may be abrupt. This problem can be solved using the suggestion of Rohsenow(1961) that considers the heat flux divided into two parts:  where the first term refers to the convection in the absence of bubbles and the second is the heat transfer only affected by the bubble movement, without convection. In the present work the first term is calculated using href obtained from Eq. A.1 and the last one by Eq. A.2. c) Critical heat flux: EPRI correlation (EPRI Report, 1983). where A and C are constants which are dependent from pressure, is the local flow quality local, and , the inlet flow quality; is the local heat flux and q"CHF , the critical heat flux. d) Transition boiling: Bjornard~s correlation (BJORNARD,1977).
The propagation of the drying and rewetting fronts is represented in Figure 6, for several mass flow rates. For mass flow rate equal to 0.0535 m/s, the drying front has a mean velocity of 4.6 cm/s. For the rewetting one, this velocity is 2.4 cm/s. The front velocities presented in Figs.6, 7 e 8 are mean velocities which are calculated dividing the maximum distance reached by the front by the correspondent time interval. The mass flow influence on the velocities is also shown. Variations of + 20% and - 20% on the reference case ( = 0.0535 kg/s) are applied. It is noted that, as this variable rises, the drying velocity also rises from 4.6 to 4.9 cm/s, while the rewetting one goes from 2.4 to 4.4 cm/s. The CHF does not occur immediately after the power step, due to the radial and axial thermal resistances and condutances of the indirect heated rod. The time delay observed is reduced from 1.0 s to 0.8 s as the mass flow rate increases.
where q"TB is the heat flux in the transition boiling region, and q"MSFB is the heat flux at the Leidenfrost temperature (TMSFB ). e) Minimum heat flux (Leidenfrost point): BJORNARD~s correlation (1977). where THN is the homogeneous nucleation temperature. This is the temperature at which the nucleation occurs spontaneously in the liquid in the absence of preferred nucleation sites. It is function of the pressure and can be predicted using standard nucleation theory. It can be obtained by the following expression of the TRAC -PF1 handbook:  where P = 3203.6 - P, ( in psia) and THN in 0F. f) Film boiling: modified Groeneveld~s correlation (BJORNARD, (1977)). where: Y = two-phase flow factor of Miropol~skiy, x = flow quality; Prwall = Prandtl number evaluated at temperature Twall; G = coolant mass flux; De = hydraulic diameter of the channel; g = dynamic viscosity - gaseous phase; g = coolant density - gaseous phase; f = coolant density - liquid phase; kg.= thermal conductivity -gaseous phase. References BJORNARD, T. A.; GRIFFITH, P. PWR blowdown heat transfer. In: Symposium on the thermal and hydraulics aspects of nuclear reactor safety, vol.1: Light Waters Reactors, pp. 17-39, 1977. [ Links ]
 
CARLSON, R.W., "Spreading of critical heat flux region during testing for onset of critical heat flux", Ann. Nucl. Energy, vol. 6, no. 2, pp. 49-62, 1989. [ Links ] GRIFFITH, P., MOHAMED, J. A. & BROWN, D., "Dryout front modeling for rod bundles", Nucl. Engin. and Design, vol. 105, pp. 223-229, 1988. [ Links ] GUNNERSON, F. S. & YACKLE, T. R., "Quenching and rewetting of nuclear fuel rods", Nuclear Technology, vol. 54, pp. 113-117, 1981. [ Links ] LIMA, R. DE C. F. DE, "Comportamento de vareta aquecida eletricamente durante transit&#xf3;rio de fluxo critico de calor", Doctoral Thesis, Instituto de Pesquisas Energ&#xe9;ticas e Nucleares / USP , So Paulo, 1997. [ Links ] MOSAAD, M. Subcooled boiling heat transfer to flowing water in a vertical tube. Doctoral Thesis, Technischen Universitaet Berlin , 1988. [ Links ] OLEK, S., ZVIRIN, Y. & ELIAS, E., "Rewetting of rod surfaces by falling liquid film as a conjugate heat transfer problem", Int. J. Multiphase Flow, vol. 14, no. 1, pp. 13-33, 1988. [ Links ] PARAMETRIC STUDY of CHF data, volume 2. A generalized subchannel CHF correlation for PWR and BWR fuel assemblies. Prepared for Electric Power Research Institut, California, 1983. [ Links ] SILVA NETO, A. J. da, ROBERTY, N.C., CARMO, E.G.D. CRISTE - um subc&#xf3;digo para o c&#xe1;lculo da distribui&#xe7;o axial, transiente, de temperaturas no canal de um reator PWR. Internal Report PEN-132, COPPE/UFRJ, Rio de Janeiro, 1983. [ Links ] TONG, L.S. & WEISMAN, J. Thermal analysis of pressurized water reactors, American Nuclear Society, 1979. [ Links ] TRAC-PF1. An advanced best-estimate computer program for pressurized water-reactor analysis. Safety Code Development Group Energy Division. [ Links ] YU, S.K.W., FARMER, P. R. & CONEY, M.W.E., "Methods and correlations for the prediction of quenching rates on hot surfaces", Int. J. Multiphase Flow, vol. 3, pp. 415-448, 1977. [ Links ] 
In order to observe the pressure influence on the reported velocities, the same transient is then analyzed under the coolant pressure of 8.0 MPa. Its inlet temperature is now equal to 280 oC. A 41% reduction in the pressure value has a strong influence on the rewetting front velocity: an increase of 936%. Otherwise there is a little influence on the drying one: it goes from 3.8 to 3.2 cm/s. These comparisons are shown in Fig. 7. The pressure also has considerable influence on the time delay which varies from 1.9 s to 1.4 s as the pressure changes from 8.0 MPa to 13.5 MPa.
The influence of the inlet mass flow rate was also investigated for the 8.0 MPa coolant pressure. Other authors have shown that this parameter has accentuated influence on the rewetting velocity at low pressures until 6.9 MPa. The test with a lower pressure was done to validate the model. The result obtained with the present model shown in Fig. 8 and confirms the trend. Note that the inlet mass flow rate has a smaller effect in the drying front. For a 20% reduction in mass flow the rewetting front velocity is reduced in 29.5%: It varies from 24.7 to 17.4 cm/s.
Conclusions
 
The present work analyzes the front propagation velocity for the drying out and rewetting processes, during the occurrence of critical heat flux in electrically heated simulators of nuclear fuel rods, caused by a power step.
This study is very important in the simulation of nuclear power plants as well as in metallurgical problems. At the occurrence of CHF, the amount of time required to cut off the electric power used in the heating of the simulator needs to be quantified. After the p ower cut off, the surface is rewetted when the temperature of the wall is less than the critical one. The two phenomena were analyzed individually by several authors (Carlson (1989),
Olek et al. (1988), Griffith et al.(1988)) and there were no information about the amount of time available for the operation of the protection systems.
Specially, there were few informations about rewetting, studied before only for descending films at low pressures. The work here presented supplies part of this lack of information.
Due to the radial and axial thermal resistances and capacitances of the indirect heating of the rod, the critical heat flux does not occur immediately after the power step. A certain time delay is observed. This time delay is reduced by increasing the pressure or the mass flow rate. In the beginning of the occurrence of CHF, the velocity of the propagation of the drying front is very high, being reduced gradually to an approximately constant value, around 4.6 cm/s. After the power cut off, the rewetting front presents a very large velocity of propagation, which is greatly affected by the system pressure and mass flow rate. The rewetting velocities were 24.7 cm/s and 2.4 cm/s for pressures of 8.0 MPa and 13.5 MPa, respectively. For the case of pressure of 8.0 MPa, the rewetting front propagation velocities were 17.4 cm/s and 24.7 cm/s for flow rates of 0.0471 kg/s and 0.0539 kg/s, respectively.
At the spot where the occurrence of CHF first starts, it was observed that the temperature increases at a rate of 118&deg; C/s, which indicates that the wall temperature would reach its temperature limit, estimated around 1100&deg; C, in approximately 7s. This is the amount of time available to turn off the electrical power supply. This observed heating rate is much larger than the value obtain by Mosaad (1988).
Additional details of the present work can be obtained in Lima (1997).
Acknowledgements The authors thank to FACEPE ( Funda&#xe7;o de Apoio  Cincia e Tecnologia do Estado de Pernambuco) for the support given to this work.
Appendix A


WBNP-4   4.4-7 4.4.2.2.4 Surface Heat Transfer Coefficients The fuel rod surface heat transfer coefficients during subcooled forced convection and nucleate boiling are presented in Section 4.4.2.8.1. 4.4.2.2.5 Fuel Clad Temperatures The outer surface of the fuel rod at the hot spot operates at a temperature of approximately 660&deg;F for steady state operation at rated power throughout core life due to the onset of nucleate boiling. Initially (beginning-of-life), this temperature is that of the clad metal outer surface.
a) Forced convection: DITTUS - BOELTER s correlation (BJORNARD (1977)).
During operation over the life of the core, the buildup of oxides and crud on the fuel rod surface causes the clad surface temperature to increase. Allowance is made in the fuel center melt evaluation for this temperature rise. Since the thermal-hydraulic design basis limits DNB, adequate-heat transfer is provided between the fuel clad and the reactor coolant so that the core thermal output is not limited by considerations of clad temperature. 4.4.2.2.6 Treatment of Peaking Factors The total heat flux hot channel factor, FQ, is defined by the ratio of the maximum to core average heat flux and is presented in Table 4.3-2 and discussed in Section 4.3.2.2.6. This results in a peak local power of 5.52 kW/ft x FQ at full-power conditions. As described in Section 4.3.2.2.6, the peak linear power for determination of protection setpoints is 22.4 kW/ft. The center line temperature at this kW/ft must be below the UO2 melt temperature over the lifetime of the rod, including allowances for uncertainties. The fuel temperature design basis is discussed in Subsection 4.4.1.2 and results in a maximum allowable calculated centerline temperature of 4700 &deg;F. The peak linear power for prevention of centerline melt is > 22.4 kW/ft.
where k is the coolant thermal conductivity; De , the hydraulic diameter of the channel; Re, the Reynolds number; and Pr , the Prandtl one.
The centerline temperature at the peak linear power resulting from overpower transients/overpower errors (assuming a maximum overpower of 121%) is below that required to produce melting. Fuel centerline temperature at rated (100%) power and at the peak linear power for the determination of protection setpoints are presented in Table 4.4-1. 4.4.2.3 Critical Heat Flux Ratio or Departure from Nucleate Boiling Ratio and Mixing Technology The minimum DNBRs for the rated power, design overpower and anticipated transient conditions are given in Table 4.4-1. The minimum DNBR in the limiting flow channel will be downstream of the peak heat flux location (hot spot) due to the increased down stream enthalpy rise.
b) Nucleate boiling: THOM      s correlation (TONG & WEISMAN, (1979)).
where Twall = external temperature of the cladding; Tsat = saturation temperature; q"SUP = heat flux; e p = pressure.
The transition between forced convection and nucleate boiling may be abrupt. This problem can be solved using the suggestion of Rohsenow(1961) that considers the heat flux divided into two parts:
where the first term refers to the convection in the absence of bubbles and the second is the heat transfer only affected by the bubble movement, without convection. In the present work the first term is calculated using href obtained from Eq. A.1 and the last one by Eq. A.2.
c) Critical heat flux: EPRI correlation (EPRI Report, 1983).
where A and C are constants which are dependent from pressure, is the local flow quality local, and    , the inlet flow quality; is the local heat flux and q"CHF , the critical heat flux.
d) Transition boiling: Bjornard    s correlation (BJORNARD,1977).
 
where q"TB is the heat flux in the transition boiling region, and q"MSFB is the heat flux at the Leidenfrost temperature (TMSFB ).
e) Minimum heat flux (Leidenfrost point): BJORNARD s correlation (1977).
where THN is the homogeneous nucleation temperature. This is the temperature at which the nucleation occurs spontaneously in the liquid in the absence of preferred nucleation sites. It is function of the pressure and can be predicted using standard nucleation theory. It can be obtained by the following expression of the TRAC -PF1 handbook:
where P = 3203.6 - P, ( in psia) and THN in 0F.
f) Film boiling: modified Groeneveld   s correlation (BJORNARD, (1977)).
where: Y = two-phase flow factor of Miropol skiy, x = flow quality; Prwall =
Prandtl number evaluated at temperature Twall; G = coolant mass flux; De =
hydraulic diameter of the channel; g = dynamic viscosity - gaseous phase; g = coolant density - gaseous phase; f = coolant density - liquid phase; kg.= thermal conductivity -gaseous phase.
References BJORNARD, T. A.; GRIFFITH, P. PWR blowdown heat transfer. In:
Symposium on the thermal and hydraulics aspects of nuclear reactor safety, vol.1: Light Waters Reactors, pp. 17-39, 1977. [ Links ]
 
CARLSON, R.W., "Spreading of critical heat flux region during testing for onset of critical heat flux", Ann. Nucl. Energy, vol. 6, no. 2, pp. 49-62, 1989.
[ Links ]
GRIFFITH, P., MOHAMED, J. A. & BROWN, D., "Dryout front modeling for rod bundles", Nucl. Engin. and Design, vol. 105, pp. 223-229, 1988. [ Links ]
GUNNERSON, F. S. & YACKLE, T. R., "Quenching and rewetting of nuclear fuel rods", Nuclear Technology, vol. 54, pp. 113-117, 1981. [ Links ]
LIMA, R. DE C. F. DE, "Comportamento de vareta aquecida eletricamente durante transit&#xf3;rio de fluxo critico de calor", Doctoral Thesis, Instituto de Pesquisas Energ&#xe9;ticas e Nucleares / USP , So Paulo, 1997. [ Links ]
MOSAAD, M. Subcooled boiling heat transfer to flowing water in a vertical tube. Doctoral Thesis, Technischen Universitaet Berlin , 1988. [ Links ]
OLEK, S., ZVIRIN, Y. & ELIAS, E., "Rewetting of rod surfaces by falling liquid film as a conjugate heat transfer problem", Int. J. Multiphase Flow, vol. 14, no. 1, pp. 13-33, 1988. [ Links ]
PARAMETRIC STUDY of CHF data, volume 2. A generalized subchannel CHF correlation for PWR and BWR fuel assemblies. Prepared for Electric Power Research Institut, California, 1983. [ Links ]
SILVA NETO, A. J. da, ROBERTY, N.C., CARMO, E.G.D. CRISTE - um subc&#xf3;digo para o c&#xe1;lculo da distribui&#xe7;o axial, transiente, de temperaturas no canal de um reator PWR. Internal Report PEN-132, COPPE/UFRJ, Rio de Janeiro, 1983. [ Links ]
TONG, L.S. & WEISMAN, J. Thermal analysis of pressurized water reactors, American Nuclear Society, 1979. [ Links ]
TRAC-PF1. An advanced best-estimate computer program for pressurized water-reactor analysis. Safety Code Development Group Energy Division. [
Links ]
YU, S.K.W., FARMER, P. R. & CONEY, M.W.E., "Methods and correlations for the prediction of quenching rates on hot surfaces", Int. J. Multiphase Flow, vol. 3, pp. 415-448, 1977. [ Links ]
 
WBNP-4 4.4.2.2.4       Surface Heat Transfer Coefficients The fuel rod surface heat transfer coefficients during subcooled forced convection and nucleate boiling are presented in Section 4.4.2.8.1.
4.4.2.2.5       Fuel Clad Temperatures The outer surface of the fuel rod at the hot spot operates at a temperature of approximately 660&deg;F for steady state operation at rated power throughout core life due to the onset of nucleate boiling.
Initially (beginning-of-life), this temperature is that of the clad metal outer surface.
During operation over the life of the core, the buildup of oxides and crud on the fuel rod surface causes the clad surface temperature to increase. Allowance is made in the fuel center melt evaluation for this temperature rise. Since the thermal-hydraulic design basis limits DNB, adequate-heat transfer is provided between the fuel clad and the reactor coolant so that the core thermal output is not limited by considerations of clad temperature.
4.4.2.2.6       Treatment of Peaking Factors The total heat flux hot channel factor, FQ, is defined by the ratio of the maximum to core average heat flux and is presented in Table 4.3-2 and discussed in Section 4.3.2.2.6.
This results in a peak local power of 5.52 kW/ft x FQ at full-power conditions. As described in Section 4.3.2.2.6, the peak linear power for determination of protection setpoints is 22.4 kW/ft.
The center line temperature at this kW/ft must be below the UO2 melt temperature over the lifetime of the rod, including allowances for uncertainties. The fuel temperature design basis is discussed in Subsection 4.4.1.2 and results in a maximum allowable calculated centerline temperature of 4700 &deg;F. The peak linear power for prevention of centerline melt is > 22.4 kW/ft.
The centerline temperature at the peak linear power resulting from overpower transients/overpower errors (assuming a maximum overpower of 121%) is below that required to produce melting. Fuel centerline temperature at rated (100%) power and at the peak linear power for the determination of protection setpoints are presented in Table 4.4-1.
4.4.2.3         Critical Heat Flux Ratio or Departure from Nucleate Boiling Ratio and Mixing Technology The minimum DNBRs for the rated power, design overpower and anticipated transient conditions are given in Table 4.4-1. The minimum DNBR in the limiting flow channel will be downstream of the peak heat flux location (hot spot) due to the increased down stream enthalpy rise.
DNBRs are calculated by using the correlation and definitions described in the following Sections 4.4.2.3.1 and 4.4.2.3.2. The VIPRE-01 computer code (discussed in Section 4.4.3.4.1) is used to determine the flow distribution in the core and the local conditions in the hot channel for use in the DNB correlation. The use of hot channel factors is discussed in Section 4.4.3.2.1 (nuclear hot channel factors) and in Section 4.4.2.3.4 (engineering hot channel factors).
DNBRs are calculated by using the correlation and definitions described in the following Sections 4.4.2.3.1 and 4.4.2.3.2. The VIPRE-01 computer code (discussed in Section 4.4.3.4.1) is used to determine the flow distribution in the core and the local conditions in the hot channel for use in the DNB correlation. The use of hot channel factors is discussed in Section 4.4.3.2.1 (nuclear hot channel factors) and in Section 4.4.2.3.4 (engineering hot channel factors).
WBN Unit 1 Nuclear Power Generation/ATWS 1-FR-S.1 Rev. 0001   Step   Action/Expected Response   Response Not Obtained   Page 10 of 16  19. CHECK Incore T/Cs less than 1200&deg;F. IF Incore T/Cs are greater than 1200&deg;F AND rising, THEN ** GO TO 1-SACRG-1, Severe Accident Control Room Guideline Initial Response. 20. CHECK reactor subcritical: a. Power range channels less than 5%. b. Intermediate range startup rate NEGATIVE. CONTINUE to borate. IF boration is NOT available, THEN ALLOW RCS to heat up to insert negative reactivity from temperature coefficients. IF red OR orange condition exists on other Status Trees, THEN PERFORM actions of other FR Procedures which do not cool down or otherwise add positive reactivity to the core. ** GO TO Step 4. 21. TERMINATE emergency boration: a. PLACE BA transfer pumps in SLOW speed. b. CLOSE emergency borate valve 1-FCV-62-138. c. IF alternate boration opened, THEN Locally CLOSE 1-ISV-62-929.  
4.4-7
 
WBN               Nuclear Power Generation/ATWS             1-FR-S.1 Unit 1                                                      Rev. 0001 Step       Action/Expected Response                 Response Not Obtained
: 19.         CHECK Incore T/Cs less than               IF Incore T/Cs are greater than 1200&deg;F 1200&deg;F.                                  AND rising, THEN If this transition is missed,                        ** GO TO 1-SACRG-1, Severe then the following step will                        Accident Control Room Guideline Initial Response.
be used.
: 20.         CHECK reactor subcritical:                CONTINUE to borate.
: a. Power range channels less than       IF boration is NOT available, 5%.                                  THEN ALLOW RCS to heat up to insert
: b. Intermediate range startup rate negative reactivity from temperature NEGATIVE.
coefficients.
If one misses the                                  IF red OR orange condition exists on transition to 1-SACRG-1,                            other Status Trees, then one will use 1-FR-                            THEN C.1.                                                PERFORM actions of other FR Procedures which do not cool down or otherwise add positive reactivity to the core.
                                                      ** GO TO Step 4.
: 21.         TERMINATE emergency boration:
: a. PLACE BA transfer pumps in SLOW speed.
: b. CLOSE emergency borate valve 1-FCV-62-138.
: c. IF alternate boration opened, THEN Locally CLOSE 1-ISV-62-929.
Page 10 of 16


WBN Unit 1 Status Trees FR-0 Rev. 0014 Attachment 1 (Page 2 of 8) Monitoring Critical Safety Functions Page 5 of 11  CORE COOLING FR-C Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 6 of 16  C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)] Some examples of SRO exam items for this topic include:
WBN                         Status Trees                 FR-0 Unit 1                                                    Rev. 0014 If the transition is          Attachment 1 missed at step 19,            (Page 2 of 8) this first decisionMonitoring Critical Safety Functions block will indicate a red path.                  CORE COOLING FR-C Page 5 of 11
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]
Some examples of SRO exam items for this topic include:
* 10 CFR 50.59 screening and evaluation processes.
* 10 CFR 50.59 screening and evaluation processes.
* Administrative processes for temporary modifications.
* Administrative processes for temporary modifications.
* Administrative processes for disabling annunciators.
* Administrative processes for disabling annunciators.
* Administrative processes for the installation of temporary instrumentation.
* Administrative processes for the installation of temporary instrumentation.
* Processes for changing the plant or plant procedures. Section IV provides an example of a satisfactory SRO-only question related to this topic. D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.       [10 CFR 55.43(b)(4)] Some examples of SRO exam items for this topic include:
* Processes for changing the plant or plant procedures.
Section IV provides an example of a satisfactory SRO-only question related to this topic.
D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
[10 CFR 55.43(b)(4)]
Some examples of SRO exam items for this topic include:
* Process for gaseous/liquid release approvals, i.e., release permits.
* Process for gaseous/liquid release approvals, i.e., release permits.
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures. Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits. SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc. E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedure's overall mitigative strategy or purpose.  
Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.
SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.
E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.
The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:
Page 6 of 16


The applicant's knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedure's content is required to correctly answer the written test item, for example: 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 7 of 16
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
* Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
* Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
SRO-only knowledge should not be claimed for questions that can be answered solely using "systems knowledge"; e.g.:
SRO-only knowledge should not be claimed for questions that can be answered solely using systems knowledge; e.g.:
* how the system works.
* how the system works.
* system flow path.
* system flow path.
* component locations, etc. SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:
* component locations, etc.
SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:
* the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
* the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
* any AOP entry condition.
* any AOP entry condition.
* plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
* plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
* immediate operator actions of a procedure. Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.        
* immediate operator actions of a procedure.
Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.
Page 7 of 16
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)
(Assessment and selection of procedures)
Can the question be answered solely by knowing systems knowledge, i.e., how the system works,      Yes RO question flowpath, logic, component location?
No Can the question be answered solely by knowing immediate operator actions?                            Yes    RO question No Can the question be answered solely by knowing entry conditions for AOPs or plant parameters          Yes RO question that require direct entry to major EOPs?
No Can the question be answered solely by knowing the purpose, overall sequence of events, or            Yes RO question overall mitigative strategy of a procedure?
No Does the question require one or more of the following?
* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps                                            Yes        SRO-only
* Knowledge of diagnostic steps and decision points in the              question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 8 of 16   Figure 2:  Screening for SRO-only linked to 10 CFR 55.43(b)(5) (Assessment and selection of procedures)      
92.
Given the following conditions:
    -  Unit 1 is at 100% power.
    -  Maintenance personnel commence Appendix J, Penetration X-80 LLRT of 1-SI-30-701.
    -  In accordance with Appendix J, 1-FCV-30-37 and 1-FCV-30-40 are SHUT.
    - Containment pressure approaches T/S LCO 3.6.4 limits.
Flow diagram of X-80 Which ONE of the following describes an action that will maintain containment pressure within the T/S LCO 3.6.4 limits?
____(1)____ AND in accordance with the ODCM, would be authorized by a _____(2)_____ release permit.
NOTE:        1-SI-30-701, Containment Isolation Valve Local Leak Rate Test Purge Air 1-FCV-30-37, LWR CNTMT PURGE EXH PRESS RLF 1-FCV-30-40, LWR CNTMT PURGE EXH PRESS RLF SOI-30.02, Containment Purge System SOI-65.02, Emergency Gas Treatment System T/S LCO 3.6.4, Containment Pressure A.    (1)  EGTS would be started in accordance with SOI-65.02 (2)  weekly periodic B.    (1)  EGTS would be started in accordance with SOI-65.02 (2)  conditional C.    (1)  Containment purge would be started in accordance with SOI-30.02 (2)   weekly periodic D.    (1)   Containment purge would be started in accordance with SOI-30.02 (2)   conditional


Can the question be answered solely by knowing "systems knowledge", i.e., how the system works, flowpath, logic, component location? RO question YesNoCan the question be answered solely by knowing immediate operator actions? YesCan the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?YesNoDoes the question require one or more of the following?
CORRECT ANSWER:                                                              D DISTRACTOR ANALYSIS:
* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
A. Incorrect: As mentioned in the (D) distractor it is incorrect and plausible that the EGTS system be used for containment pressure control. Of note is the fact that the permitting requirements for an EGTS subsystem operation are component to 1-ODI-90-26.
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps
B. Incorrect: While the need to obtain a prior to release or conditional permit is correct: the utilization of the EGTS system is not correct. As previously discussed, EGTS would assist in controlling containment pressure ONLY if the two dampers 1-FCV-30-37 and 1-FCV-30-40 were OPEN. A manual start of the EGTS subsystem (with no containment isolation phase A signal present) would assist in maintaining containment pressure because during such scenario, the containment and the annulus volumes would be cross-connected.
* Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
Therefore, choosing such distractor is plausible but incorrect as the containment pressure relief valves were closed in support of the LLRT.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?NoNoYesQuestion might not be linked to 10 CFR 55.43(b)(5) for SRO-only NoRO question RO question RO question 
C. Incorrect: Again, it is correct that the containment purge would be utilized for containment pressure control. It is not correct however, that the weekly permit resultant from 1-ODI-90-26 would authorize this discharge. It is plausible to believe this as the purge does discharge to the Units shield building stack and as 1-ODI-90-26 samples the shield building exhaust the implication is present that such procedure would enable the containment purge to be accomplished.
: 92. Given the following conditions:   - Unit 1 is at 100% power.
D. Correct: As seen in system description WBN-SDD-N3-30RB-4002, Reactor Building Ventilation System: The containment venting, for continuous pressure relief, is performed during modes 1-5, by opening the containment isolation (CI) valves FCV-30-40 and -37. This allows continuous venting of containment air into the Annulus through one of the Containment Vent Air Cleanup Units. This is also seen on flow print 1-47W866-1. Therefore, if the two FCVs mentioned are shut, then the pressure inside of containment will uncontrollably rise. As seen in system description N3-65-4001, Emergency Gas Treatment System: the EGTS establishes and keeps the annulus at a negative pressure and captures containment out-leakage. The EGTS is placed into service after a containment isolation (phase A) is received.
- Maintenance personnel commence Appendix J, Penetration X-80 LLRT of 1-SI-30-701. - In accordance with Appendix J, 1-FCV-30-37 and 1-FCV-30-40 are SHUT. - Containment pressure approaches T/S LCO 3.6.4 limits. Flow diagram of X-80  Which ONE of the following describes an action that will maintain containment pressure within the T/S LCO 3.6.4 limits?
Furthermore, the flowpath through the EGTS subsystem shows that air is drawn from the annulus, processed through filter banks and then discharged with some flow being returned to the annulus and the remainder being sent out of the Units shield building exhaust stack.
This division of discharge maintains the annulus at a slight vacuum (relative to the isolated containment vessel). One may see that the EGTS subsystem is not designed to maintain the containment pressure within any bounds; it simply addresses any leakage which emanates from the isolated containment. Therefore, it is correct that initiating a containment purge will be the sole viable option for maintaining containment pressure within the limits of the Technical


____(1)____ AND in accordance with the ODCM, would be authorized by a _____(2)_____ release permit. A. (1) EGTS would be started in accordance with SOI-65.02 (2) weekly periodic B. (1) EGTS would be started in accordance with SOI-65.02 (2) conditional C. (1) Containment purge would be started in accordance with SOI-30.02 (2) weekly periodic D. (1) Containment purge would be started in accordance with SOI-30.02 (2) conditional  NOTE: 1-SI-30-701, Containment Isolation Valve Local Leak Rate Test Purge Air  1-FCV-30-37, LWR CNTMT PURGE EXH PRESS RLF  1-FCV-30-40, LWR CNTMT PURGE EXH PRESS RLF  SOI-30.02, Containment Purge System SOI-65.02, Emergency Gas Treatment System  T/S LCO 3.6.4, Containment Pressure CORRECT ANSWER:D DISTRACTOR ANALYSIS:  A. Incorrect: As mentioned in the (D) distractor it is incorrect and plausible that the EGTS system be used for containment pressure control. Of note is the fact that the permitting requirements for an EGTS subsystem operation are component to 1-ODI-90-26. B. Incorrect: While the need to obtain a "prior to release" or conditional permit is correct: the utilization of the EGTS system is not correct. As previously discussed, EGTS would assist in controlling containment pressure ONLY if the two dampers 1-FCV-30-37 and 1-FCV-30-40 were OPEN. A manual start of the EGTS subsystem (with no containment isolation phase A signal present) would assist in maintaining containment pressure because during such scenario, the containment and the annulus volumes would be cross-connected. Therefore, choosing such distractor is plausible but incorrect as the containment pressure relief valves were closed in support of the LLRT. C. Incorrect: Again, it is correct that the containment purge would be utilized for containment pressure control. It is not correct however, that the weekly permit resultant from 1-ODI-90-26 would authorize this discharge. It is plausible to believe this as the purge does discharge to the Unit's shield building stack and as 1-ODI-90-26 samples the shield building exhaust the implication is present that such procedure would enable the containment purge to be accomplished. D. Correct: As seen in system description WBN-SDD-N3-30RB-4002, "Reactor Building Ventilation System:"  "The containment venting, for continuous pressure relief, is performed during modes 1-5, by opening the containment isolation (CI) valves FCV-30-40 and -37. This allows continuous venting of containment air into the Annulus through one of the Containment Vent Air Cleanup Units."  This is also seen on flow print 1-47W866-1. Therefore, if the two FCVs mentioned are shut, then the pressure inside of containment will uncontrollably rise. As seen in system description N3-65-4001, "Emergency Gas Treatment System:"  the EGTS establishes and keeps the annulus at a negative pressure and captures containment out-leakage. The EGTS is placed into service after a containment isolation (phase A) is received. Furthermore, the flowpath through the EGTS subsystem shows that air is drawn from the annulus, processed through filter banks and then discharged with some flow being returned to the annulus and the remainder being sent out of the Unit's shield building exhaust stack. This division of discharge maintains the annulus at a slight vacuum (relative to the isolated containment vessel). One may see that the EGTS subsystem is not designed to maintain the containment pressure within any bounds; it simply addresses any leakage which emanates from the isolated containment. Therefore, it is correct that initiating a containment purge will be the sole viable option for maintaining containment pressure within the limits of the Technical Specifications.
Specifications.
As seen in Table 2.2-2 of the Offsite Dose Calculation Manual (ODCM), a containment purge requires that both the minimum sampling and analysis frequencies are "P each purge."  As seen in table 3.1 of the ODCM, this annotation indicates that a sample and analysis of the containment atmosphere must be "completed prior to each release."  As seen on print 1-47W866-1 (or on the simplified ICS screen shot), a containment purge is discharged to the Unit's shield building stack. Aforementioned was the fact that the EGTS subsystem also discharges (at least for a portion of its flow) to the shield building stack. Such a shield building stack discharge sampled and analyzed on a "W" frequency. This is contained again in table 2.2-2. A "W" frequency is "at least once per 7 days."  At WBN, the requirements of the ODCM are in part implemented by the Offsite Dose Instructions (ODIs). Of import to this question are two ODIs. 1-ODI-90-15, "Containment Purge Release" satisfies the requirements applicable to the containment purge in table 2.2-2. 1-ODI-90-26, "Weekly Sampling Of Unit 1 Shield Building Exhaust" addresses the ODCM compliance of the other effluents discharging through the Unit's shield building stack. The outputs of these ODIs are the release permits. Therefore, it is correct that before containment purge is to be initiated, that a release permit be authorized.
As seen in Table 2.2-2 of the Offsite Dose Calculation Manual (ODCM), a containment purge requires that both the minimum sampling and analysis frequencies are P each purge. As seen in table 3.1 of the ODCM, this annotation indicates that a sample and analysis of the containment atmosphere must be completed prior to each release. As seen on print 1-47W866-1 (or on the simplified ICS screen shot), a containment purge is discharged to the Units shield building stack. Aforementioned was the fact that the EGTS subsystem also discharges (at least for a portion of its flow) to the shield building stack. Such a shield building stack discharge sampled and analyzed on a W frequency. This is contained again in table 2.2-2. A W frequency is at least once per 7 days. At WBN, the requirements of the ODCM are in part implemented by the Offsite Dose Instructions (ODIs). Of import to this question are two ODIs. 1-ODI-90-15, Containment Purge Release satisfies the requirements applicable to the containment purge in table 2.2-2. 1-ODI-90-26, Weekly Sampling Of Unit 1 Shield Building Exhaust addresses the ODCM compliance of the other effluents discharging through the Units shield building stack. The outputs of these ODIs are the release permits. Therefore, it is correct that before containment purge is to be initiated, that a release permit be authorized.
Question Number: 92  Tier:  2 Group:  2 K/A: 029 Containment Purge System A2 Ability to (a) predict the impacts of the following malfunctions or operations on the Containment Purge System; and (b) based on those predictions, use procedures to Correct: control, or mitigate the consequences of those malfunctions or operations:
A2.01 Maintenance or other activity taking place inside containment Importance Rating: 2.9  3.6  10 CFR Part 55: (CFR: 41.5 / 43.5 / 45.3 / 45.13)


10CFR55.43.b: 10 CFR 55.43(b)(2) and 10 CFR 55.43(b)(5) K/A Match: K/A is matched because the applicant is required to predict the impact that the performance of Appendix J of 1-SI-30-701 (a maintenance activity which is conducted inside of the containment) has upon the containment purge system. Namely, the applicant must identify that it would require that containment purge be started if containment pressure control became needed. Subsequently, the applicant must use the information in the ODCM to correctly select the required permitting to conduct the purge. Technical  
Question Number:        92 Tier:    2  Group:        2 K/A:    029 Containment Purge System A2 Ability to (a) predict the impacts of the following malfunctions or operations on the Containment Purge System; and (b) based on those predictions, use procedures to Correct: control, or mitigate the consequences of those malfunctions or operations:
A2.01 Maintenance or other activity taking place inside containment Importance Rating:      2.9 3.6 10 CFR Part 55:        (CFR: 41.5 / 43.5 / 45.3 / 45.13) 10CFR55.43.b:         10 CFR 55.43(b)(2) and 10 CFR 55.43(b)(5)
K/A Match:   K/A is matched because the applicant is required to predict the impact that the performance of Appendix J of 1-SI-30-701 (a maintenance activity which is conducted inside of the containment) has upon the containment purge system.
Namely, the applicant must identify that it would require that containment purge be started if containment pressure control became needed. Subsequently, the applicant must use the information in the ODCM to correctly select the required permitting to conduct the purge.
Technical  


==Reference:==
==Reference:==
Offsite Dose Calculation Manual (ODCM) 1-ODI-90-15, Containment Purge Release 1-ODI-90-26, Weekly Sampling Of Unit 1 Shield Building Exhaust 1-SI-30-701, Cntmt Isol Vlv Local LR Test Purge Air 1-47W866-1 ICS screenshot of EFF1 screen. Proposed references to be provided: None  Learning Objective:   3-OT-SYS065A 9. Given plant conditions, IDENTIFY the applicable EGTS System limits and precautions related to the following: b. SOI-65.02 Emergency Gas Treatment System  
Offsite Dose Calculation Manual (ODCM) 1-ODI-90-15, Containment Purge Release 1-ODI-90-26, Weekly Sampling Of Unit 1 Shield Building Exhaust 1-SI-30-701, Cntmt Isol Vlv Local LR Test Purge Air 1-47W866-1 ICS screenshot of EFF1 screen.
: 12. DESCRIBE the following aspects of TS and TRs
Proposed references to     None be provided:
: b. The Limiting Conditions for Operation, Applicability, and Bases. Cognitive Level:     Higher X Lower     Question Source:     New X Modified Bank   Bank     Question History: New question for the 2015-301 NRC SRO Exam Comments:
Learning Objective:         3-OT-SYS065A
WBN System Description Document REACTOR BUILDING VENTILATION SYSTEM WBN-SDD-N3-30RB-4002 Rev. 0024 Page 50 of 96 3.1.3 Containment Air Return System (continued)     Ductwork associated with the fans consists of hydrogen collectors from the reactor cavity, the containment dome, shared collection headers from the lower compartment, the pressurizer compartment, and the steam generator compartments. 3.1.4 Containment Vent System The containment venting, for continuous pressure relief, is performed during modes 1-5, by opening the containment isolation (CI) valves FCV-30-40 and -37. This allows continuous venting of containment air into the Annulus through one of the Containment Vent Air Cleanup Units (CVACU)s, which are equipped with HEPA and charcoal filters. The airflow from containment into the Annulus is provided by the motive force of the differential pressure between the containment and the Annulus. This air mixes with the Annulus atmosphere before the AVC fan discharges it into the AB exhaust stack via the suction-side duct of the AB FHA exhaust fans. As an alternate to using the normal vent pathway, for containment pressure relief, either the pair of lower compartment purge lines (one supply and one exhaust), or one of the two pairs of upper compartment purge lines (one supply and one exhaust) may be used. The use of these alternate lines may require re-balancing of the supply duct airflow, as needed, to preclude a containment pressure rise. When an upper, or the lower, compartment purge line is used, the Containment Vent System must be isolated. The Containment Vent System shall be isolated, during mode 6, by closing the valves FCV-30-40 and FCV-30-37 (Refer to subSection 4.20). 3.2 Component Description 3.2.1 Major Component Description Note: The following is vendor data which describe the performance characteristics for major system components. For more detailed information and component requirements, the appropriate contract should be referenced. The information included in this section shall be updated upon any modification, addition, or replacement of existing equipment. The data represent the manufacturers' rated capacities and not to be construed as required design values. Refer to Section 3.1 and Table 9.6 for design values. A. Purge Supply Fans TVA Contract No. - 76K35-83246-1 Manufacturer - H. K. Porter Company, Incorporated Capacity   - 14,000 cfm at 9.5" Static Pressure Type   - Belt-Driven Centrifugal Motor   - 50 hp Seismic   - Category I B. Purge Exhaust Fans TVA Contract No. - 76K35-83246-1 Manufacturer - H. K. Porter Company, Incorporated
: 9. Given plant conditions, IDENTIFY the applicable EGTS System limits and precautions related to the following:
: b. SOI-65.02 Emergency Gas Treatment System
: 12. DESCRIBE the following aspects of TS and TRs
: b. The Limiting Conditions for Operation, Applicability, and Bases.
Cognitive Level:
Higher               X Lower Question Source:
New                 X Modified Bank Bank Question History:           New question for the 2015-301 NRC SRO Exam Comments:
 
WBN System           REACTOR BUILDING VENTILATION                 WBN-SDD-N3-30RB-4002 Description                          SYSTEM                        Rev. 0024 Document                                                          Page 50 of 96 3.1.3   Containment Air Return System (continued)
Ductwork associated with the fans consists of hydrogen collectors from the reactor cavity, the containment dome, shared collection headers from the lower compartment, the pressurizer compartment, and the steam generator compartments.
3.1.4   Containment Vent System The containment venting, for continuous pressure relief, is performed during modes 1-5, by opening the containment isolation (CI) valves FCV-30-40 and -37. This allows continuous venting of containment air into the Annulus through one of the Containment Vent Air Cleanup Units (CVACU)s, which are equipped with HEPA and charcoal filters. The airflow from containment into the Annulus is provided by the motive force of the differential pressure between the containment and the Annulus. This air mixes with the Annulus atmosphere before the AVC fan discharges it into the AB exhaust stack via the suction-side duct of the AB FHA exhaust fans. As an alternate to using the normal vent pathway, for containment pressure relief, either the pair of lower compartment purge lines (one supply and one exhaust), or one of the two pairs of upper compartment purge lines (one supply and one exhaust) may be used. The use of these alternate lines may require re-balancing of the supply duct airflow, as needed, to preclude a containment pressure rise. When an upper, or the lower, compartment purge line is used, the Containment Vent System must be isolated.
The Containment Vent System shall be isolated, during mode 6, by closing the valves FCV-30-40 and FCV-30-37 (Refer to subSection 4.20).
3.2     Component Description 3.2.1   Major Component Description Note:   The following is vendor data which describe the performance characteristics for major system components. For more detailed information and component requirements, the appropriate contract should be referenced. The information included in this section shall be updated upon any modification, addition, or replacement of existing equipment. The data represent the manufacturers' rated capacities and not to be construed as required design values. Refer to Section 3.1 and Table 9.6 for design values.
A. Purge Supply Fans TVA Contract No.     -   76K35-83246-1 Manufacturer         -   H. K. Porter Company, Incorporated Capacity             -   14,000 cfm at 9.5" Static Pressure Type                 -   Belt-Driven Centrifugal Motor                 -   50 hp Seismic               -   Category I B. Purge Exhaust Fans TVA Contract No.     -   76K35-83246-1 Manufacturer         -   H. K. Porter Company, Incorporated


NPG System Description Document EMERGENCY GAS TREATMENT SYSTEM N3-65-4001 Rev. 0010 Page 26 of 52   3.1.2 Air Cleanup Unit (ACU) Subsystem The ACU subsystem is an ESF with two independent, 100% capacity trains. Each train consists of an exhaust fan, a HEPA-charcoal filter assembly, isolation valves, associated dampers and ductwork, and instruments and controls. The ACU fans are located in the auxiliary building EGTS room adjacent to the Unit 2 shield building on El 757.0. The ACU fan design flow rate and the annulus negative pressure to be maintained are shown in Table 7. The ACU intake is centrally located within the annulus above the steel containment dome. The intakes and ducting used to bring the air to ACU subsystem are shared with the AVCS. The ACU subsystem starts automatically when a CIA is received. It provides two capabilities needed during a LOCA. One of these is the capability to reduce out-leakage of radioactive material from the shield building to within the guideline limits of Ref. 7.5.1. This is accomplished by establishing and keeping the annulus at a negative pressure (Ref. 7.4.3). The second capability is to capture containment out-leakage and process it through a series of HEPA and charcoal filters before release to the atmosphere. The ACU housing contains the following components listed in order: a moisture separator, a relative humidity heater, a prefilter, HEPA filter, two charcoal filter beds in series, and an after HEPA filter. An exhaust fan is provided downstream of each ACU housing. The air flow network can be aligned to exhaust annulus air through either EGTS filter train. This is accomplished by closing the AVCS isolation valves and opening the ACU subsystem valves. See Table 2 for a listing of valves and the valve alignment during the ACU operation. After the air cleanup subsystem has established the required annulus pressure, a maximum of 250 cfm of air is released through the shield building exhaust vent for a postulated single failure of one EGTS train or a maximum of 957 cfm for a postulated single failure of a control loop associated with one train of PCOs. (Ref 7.4.4). The remaining flow is recirculated in the annulus in a manner that promotes mixing, dilution, and holdup of the containment out-leakage. The recirculated air flow is discharged from a manifold extending completely around the bottom of the annulus. There are 23 ports in the manifold with a rated flow of 174 cfm each (Refs. 7.4.1 and 7.1.6). The vertical separation between the exhaust and the discharge ports is 168'-9". After the air has been processed, the airflow network directs the air to redundant damper controlled flow dividers in the annulus. At this point, the flow network contains two airflow paths leading to the unit's shield building exhaust vent (either 1-PCO-65-80 and 1-PCV-65-81 or 1-PCO-65-82 and 1-PCV-65-83) and two airflow paths to the annulus manifold (either 1-PCO-65-88 and 1-PCV-65-86 or 1-PCO-65-89 and 1-PCV-65-87).
NPG System         EMERGENCY GAS TREATMENT SYSTEM                   N3-65-4001 Description                                                          Rev. 0010 Document                                                            Page 26 of 52 3.1.2   Air Cleanup Unit (ACU) Subsystem The ACU subsystem is an ESF with two independent, 100% capacity trains. Each train consists of an exhaust fan, a HEPA-charcoal filter assembly, isolation valves, associated dampers and ductwork, and instruments and controls. The ACU fans are located in the auxiliary building EGTS room adjacent to the Unit 2 shield building on El 757.0. The ACU fan design flow rate and the annulus negative pressure to be maintained are shown in Table 7. The ACU intake is centrally located within the annulus above the steel containment dome. The intakes and ducting used to bring the air to ACU subsystem are shared with the AVCS.
The ACU subsystem starts automatically when a CIA is received. It provides two capabilities needed during a LOCA. One of these is the capability to reduce out-leakage of radioactive material from the shield building to within the guideline limits of Ref. 7.5.1. This is accomplished by establishing and keeping the annulus at a negative pressure (Ref. 7.4.3). The second capability is to capture containment out-leakage and process it through a series of HEPA and charcoal filters before release to the atmosphere. The ACU housing contains the following components listed in order: a moisture separator, a relative humidity heater, a prefilter, HEPA filter, two charcoal filter beds in series, and an after HEPA filter. An exhaust fan is provided downstream of each ACU housing. The air flow network can be aligned to exhaust annulus air through either EGTS filter train. This is accomplished by closing the AVCS isolation valves and opening the ACU subsystem valves. See Table 2 for a listing of valves and the valve alignment during the ACU operation. After the air cleanup subsystem has established the required annulus pressure, a maximum of 250 cfm of air is released through the shield building exhaust vent for a postulated single failure of one EGTS train or a maximum of 957 cfm for a postulated single failure of a control loop associated with one train of PCOs. (Ref 7.4.4). The remaining flow is recirculated in the annulus in a manner that promotes mixing, dilution, and holdup of the containment out-leakage. The recirculated air flow is discharged from a manifold extending completely around the bottom of the annulus. There are 23 ports in the manifold with a rated flow of 174 cfm each (Refs. 7.4.1 and 7.1.6). The vertical separation between the exhaust and the discharge ports is 168'-9".
After the air has been processed, the airflow network directs the air to redundant damper controlled flow dividers in the annulus. At this point, the flow network contains two airflow paths leading to the unit's shield building exhaust vent (either 1-PCO-65-80 and 1-PCV-65-81 or 1-PCO-65-82 and 1-PCV-65-83) and two airflow paths to the annulus manifold (either 1-PCO-65-88 and 1-PCV-65-86 or 1-PCO-65-89 and 1-PCV-65-87).
1-PCO-65-80, 82, 88 and 89 modulate to maintain the annulus pressure relative to the outside environment. The isolation dampers are zero leakage valves used to minimize outside air in-leakage from the shield building exhaust vent into the annulus. By varying the amount of air that is exhausted through the shield building exhaust vent, the negative annulus pressure is maintained. This pressure level is low enough so that leakage will be into the annulus from both primary containment and areas adjacent to the shield building.
1-PCO-65-80, 82, 88 and 89 modulate to maintain the annulus pressure relative to the outside environment. The isolation dampers are zero leakage valves used to minimize outside air in-leakage from the shield building exhaust vent into the annulus. By varying the amount of air that is exhausted through the shield building exhaust vent, the negative annulus pressure is maintained. This pressure level is low enough so that leakage will be into the annulus from both primary containment and areas adjacent to the shield building.
The pressure differentials produced by wind effects and low temperature effects (Ref. 7.4.9) are also overcome by the appropriate selection of the pressure level. The relative humidity heater and controls are arranged such that the heaters are energized whenever the EGTS ACU exhaust fans achieve a flow setpoint. The heaters de-energize when the fan flowrate is below the setpoint. Each heater is designed to maintain an air stream relative humidity of 70% before it is routed through the ACU filters in accordance with the requirements of Ref. 7.5.5.
The pressure differentials produced by wind effects and low temperature effects (Ref. 7.4.9) are also overcome by the appropriate selection of the pressure level.
NPG System Description Document EMERGENCY GAS TREATMENT SYSTEM N3-65-4001 Rev. 0010 Page 27 of 52 3.1.2 Air Cleanup Unit (ACU) Subsystem (continued)     Another feature incorporated into the ACU subsystem is the ability to cool the filters and adsorbers and remove radioactive decay heat in an inactive ACU containing radioactive material. This is accomplished with two crossover flow ducts that draw air at a minimum of 200 cfm through the active ACU from the discharge of the inactive ACU (Ref. 7.4.2). This flow rate is sufficient to limit the temperature rise in the inactive ACU to less than 75&deg;F when even it is fully loaded (Ref. 7.4.2). Two butterfly valves are utilized in the crossover path to assure isolation. The isolation valves are opened automatically when the valve control switch is in P-AUTO position and one ACU fan is operating and the other ACU fan is idle. These valves are normally closed and require operator action to be positioned in P-AUTO after an accident (see Section 4.2). However, the suction valve from the affected annulus to the inactive ACU must be opened by operator's action. Temperature rise is recorded in the MCR. The two ACUs in the subsystem have steel housings. The housings incorporate a quench-type water spray and drain system for flooding the charcoal filters in case of fire. (Ref. 7.2.23). The EGTS must start within 30 seconds upon receiving a CIA signal (Refs. 7.4.4 and 7.4.22). The purge air valves (FCV-30-2, -5, -12, -54, -61, and -62) must close to meet this requirement (Ref. 7.4.22) 3.2 Component Description 3.2.1 Major Component Description EGTS related components are in Ref. 7.2.26 and 7.1.10. The vendor data which describe basic design and performance characteristics for major system components are shown in Table 8. More detailed information and component requirements may be found in the contract drawing files. The data represents the manufacturer's rated capacities and should not to be construed as required design values. 3.2.2 Active Components Listing An active valves and dampers list is included in Table 3. 3.3 Instrumentation and Controls A detailed description of the EGTS electrical controls and logic can be found in Ref. 7.1.2. and 7.1.3. Operational limits, analytical limits, and safety limits, for instruments, as applicable, have been determined in Ref. 7.4.3. These limits have been used to establish instrument setpoints under all normal and LOCA conditions. The resulting setpoints are tabulated in the I-Tabs, 47B601-65 series drawings. (Ref. 7.1.8) 3.3.1 Instrumentation This section describes the instruments used to sense, indicate, and record flow, temperature, and pressure. Table 4 shows panel numbers for each instrument in both subsystems.
The relative humidity heater and controls are arranged such that the heaters are energized whenever the EGTS ACU exhaust fans achieve a flow setpoint. The heaters de-energize when the fan flowrate is below the setpoint. Each heater is designed to maintain an air stream relative humidity of 70% before it is routed through the ACU filters in accordance with the requirements of Ref. 7.5.5.
WBN 0  OFFSITE DOSE CALCULATION MANUAL (ODCM) Revision 25 Page 27 of 195     Table 2.2-2-RADIOACTIVE GASEOUS WASTE MONITORING SAMPLING AND ANALYSIS PROGRAM* (Page 1 of 3) Gaseous Release Type Minimum Sampling Frequency Analysis Frequency Type of  Activity Analysis Lower Limit of Detection (LLD) (Ci/ml)1 A. Waste Gas     Decay Tank P Each Tank Grab Sample P Each Tank Noble Gases2 (Gamma Emitters) 1x10-4      H-3 (oxide) 1x10-6 B. Containment      PURGE3  P8 Each PURGE Grab Sample P Each Purge Noble Gases2 (Gamma Emitters) 1x10-4     C. Incore Instrument     Room PURGE3 Each PURGE9 Grab Sample Each Purge Noble Gases2 (Gamma Emitters) 1x10-4      D. Requirement Deleted     E. Auxiliary Building Exh.3,10 F. Condenser Vacuum Exh.11 G. Service Building Exh. M Grab Sample M Noble Gases2 (Gamma Emitters) 1x10-4  H. Deleted in Revision 11. I. Deleted in Revision 11. J. Deleted in Revision 11. K. Auxiliary Building Exh. L. Shield Building Exh. M. Condenser Vacuum Exh.11,12 Continuous6 Tritium Sample W H-3 (oxide) 1x10-6  Continuous6 Charcoal Sample W7 I-131 I-133 1x10-12 1x10-10 Continuous6 Particulate Sample W7 Principal Gamma Emitters2 1x10-11 Continuous6 Composite Particulate Sample M Gross Alpha 1x10-11   Q Sr-89, Sr-90 1x10-11
 
NPG System         EMERGENCY GAS TREATMENT SYSTEM                     N3-65-4001 Description                                                            Rev. 0010 Document                                                              Page 27 of 52 3.1.2   Air Cleanup Unit (ACU) Subsystem (continued)
Another feature incorporated into the ACU subsystem is the ability to cool the filters and adsorbers and remove radioactive decay heat in an inactive ACU containing radioactive material. This is accomplished with two crossover flow ducts that draw air at a minimum of 200 cfm through the active ACU from the discharge of the inactive ACU (Ref. 7.4.2). This flow rate is sufficient to limit the temperature rise in the inactive ACU to less than 75&deg;F when even it is fully loaded (Ref. 7.4.2). Two butterfly valves are utilized in the crossover path to assure isolation. The isolation valves are opened automatically when the valve control switch is in P-AUTO position and one ACU fan is operating and the other ACU fan is idle.
These valves are normally closed and require operator action to be positioned in P-AUTO after an accident (see Section 4.2). However, the suction valve from the affected annulus to the inactive ACU must be opened by operator's action. Temperature rise is recorded in the MCR.
The two ACUs in the subsystem have steel housings. The housings incorporate a quench-type water spray and drain system for flooding the charcoal filters in case of fire.
(Ref. 7.2.23).
The EGTS must start within 30 seconds upon receiving a CIA signal (Refs. 7.4.4 and 7.4.22). The purge air valves (FCV-30-2, -5, -12, -54, -61, and -62) must close to meet this requirement (Ref. 7.4.22) 3.2     Component Description 3.2.1   Major Component Description EGTS related components are in Ref. 7.2.26 and 7.1.10.
The vendor data which describe basic design and performance characteristics for major system components are shown in Table 8. More detailed information and component requirements may be found in the contract drawing files. The data represents the manufacturer's rated capacities and should not to be construed as required design values.
3.2.2   Active Components Listing An active valves and dampers list is included in Table 3.
3.3     Instrumentation and Controls A detailed description of the EGTS electrical controls and logic can be found in Ref. 7.1.2.
and 7.1.3.
Operational limits, analytical limits, and safety limits, for instruments, as applicable, have been determined in Ref. 7.4.3. These limits have been used to establish instrument setpoints under all normal and LOCA conditions. The resulting setpoints are tabulated in the I-Tabs, 47B601-65 series drawings. (Ref. 7.1.8) 3.3.1   Instrumentation This section describes the instruments used to sense, indicate, and record flow, temperature, and pressure. Table 4 shows panel numbers for each instrument in both subsystems.
 
WBN                   OFFSITE DOSE CALCULATION MANUAL                   Revision 25 0                                    (ODCM)                           Page 27 of 195 Table 2.2-2-RADIOACTIVE GASEOUS WASTE MONITORING SAMPLING AND ANALYSIS PROGRAM*
(Page 1 of 3)
Gaseous Release Type             Minimum       Analysis   Type of Activity Lower Limit of Sampling      Frequency      Analysis        Detection Frequency                                    (LLD) (Ci/ml)1 2
A. Waste Gas                               P              P      Noble Gases          1x10-4 Decay Tank                         Each Tank         Each   (Gamma Emitters)
Grab Sample        Tank H-3 (oxide)         1x10-6 2
B. Containment                           P8              P       Noble Gases 3
PURGE                            Each PURGE         Each   (Gamma Emitters)       1x10-4 Grab Sample        Purge 2
C. Incore Instrument                     Each          Each      Noble Gases          1x10-4 3
Room PURGE                          PURGE9         Purge   (Gamma Emitters)
Grab Sample D. Requirement Deleted 3,10                                                2 E. Auxiliary Building Exh.                 M              M      Noble Gases          1x10-4 F. Condenser Vacuum Exh.11         Grab Sample                (Gamma Emitters)
G. Service Building Exh.
H. Deleted in Revision 11.
I. Deleted in Revision 11.
J. Deleted in Revision 11.
K. Auxiliary Building Exh.           Continuous6        W        H-3 (oxide) 1x10-6 L. Shield Building Exh.                 Tritium M. Condenser Vacuum Exh.11,12           Sample Continuous6         W7           I-131 1x10-12 Charcoal                      I-133 Sample                                        1x10-10 Continuous6         W7     Principal Gamma 1x10-11 Particulate                   Emitters2 Sample Continuous6          M       Gross Alpha 1x10-11 Composite Particulate Sample Q       Sr-89, Sr-90 1x10-11
* See Table 3.1 (FREQUENCY NOTATION) for the surveillance frequency definitions.
* See Table 3.1 (FREQUENCY NOTATION) for the surveillance frequency definitions.
WBN 0  OFFSITE DOSE CALCULATION MANUAL (ODCM)  Revision 24 Page 57 of 195    Table 3.1 - FREQUENCY NOTATION NOTATION FREQUENCY  S  At least once per 12 hours. D  At least once per 24 hours. W  At least once per 7 days. M  At least once per 31 days. Q  At least once per 92 days. SA  At least once per 184 days. 3Q  At least once per 276 days. Y  At least once per 365 days. R  At least once per 18 months. N/A  Not applicable. P  Completed prior to each release. 


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 3 of 16  II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]: A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)] Some examples of SRO exam items for this topic include:
WBN    OFFSITE DOSE CALCULATION MANUAL          Revision 24 0                      (ODCM)                Page 57 of 195 Table 3.1 - FREQUENCY NOTATION NOTATION FREQUENCY S          At least once per 12 hours.
D          At least once per 24 hours.
W            At least once per 7 days.
M          At least once per 31 days.
Q          At least once per 92 days.
SA          At least once per 184 days.
3Q          At least once per 276 days.
Y          At least once per 365 days.
R          At least once per 18 months.
N/A                Not applicable.
P        Completed prior to each release.
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]:
A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)]
Some examples of SRO exam items for this topic include:
* Reporting requirements when the maximum licensed thermal power output is exceeded.
* Reporting requirements when the maximum licensed thermal power output is exceeded.
* Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems, fire doors, etc.
* Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems, fire doors, etc.
* The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
* The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
* National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
* National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
* Processes for TS and FSAR changes. Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic. B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)] Some examples of SRO exam items for this topic include:
* Processes for TS and FSAR changes.
Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic.
B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]
Some examples of SRO exam items for this topic include:
* Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
* Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
* Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
* Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
* Knowledge of TS bases that are required to analyze TS required actions and terminology.
* Knowledge of TS bases that are required to analyze TS required actions and terminology.
* Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).
* Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of  1 hour action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.  
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of  1 hour action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. ROs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Page 3 of 16


SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. RO's are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 4 of 16 ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)         Above this line RO knowledge Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 5 of 16 Figure 1:  Screening for SRO-only linked to 10 CFR 55.43(b)(2) (Tech Specs)       
The information               Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) required for the ODCM would be          ACTIONS from the LCO and associated applicability statements tantamount to that      (standardized TS; see example below) in the T/S which is "below the line."
RO knowledge Above this line Page 4 of 16


Can question be answered solely by knowing  1 hour TS/TRM Action? RO question YesNoCan question be answered solely by knowing the LCO/TRM information listed "above-the-line?" YesRO question NoCan question be answered solely by knowing the TS Safety Limits? YesRO question NoDoes the question involve one or more of the following for TS, TRM, or ODCM?
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)
(Tech Specs)
Can question be answered solely by knowing  1       Yes hour TS/TRM Action?                                                 RO question No Can question be answered solely by knowing the       Yes LCO/TRM information listed above-the-line?                       RO question No Can question be answered solely by knowing the       Yes TS Safety Limits?                                                   RO question No Does the question involve one or more of the following for TS, TRM, or ODCM?
* Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
* Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)
* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)                                   Yes      SRO-only
* Knowledge of TS bases that is required to analyze TS required actions and terminology SRO-only question YesNoQuestion might not be linked to 10 CFR 55.43(b)(2) for SRO-only Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 6 of 16  C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)] Some examples of SRO exam items for this topic include:
* Knowledge of TS bases that is required to analyze TS                       question required actions and terminology No The question uses Question might not be linked to               "below the line" 10 CFR 55.43(b)(2) for SRO-only               knowledge of the ODCM.
Page 5 of 16
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]
Some examples of SRO exam items for this topic include:
* 10 CFR 50.59 screening and evaluation processes.
* 10 CFR 50.59 screening and evaluation processes.
* Administrative processes for temporary modifications.
* Administrative processes for temporary modifications.
* Administrative processes for disabling annunciators.
* Administrative processes for disabling annunciators.
* Administrative processes for the installation of temporary instrumentation.
* Administrative processes for the installation of temporary instrumentation.
* Processes for changing the plant or plant procedures. Section IV provides an example of a satisfactory SRO-only question related to this topic. D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.       [10 CFR 55.43(b)(4)] Some examples of SRO exam items for this topic include:
* Processes for changing the plant or plant procedures.
Section IV provides an example of a satisfactory SRO-only question related to this topic.
D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
[10 CFR 55.43(b)(4)]
Some examples of SRO exam items for this topic include:
* Process for gaseous/liquid release approvals, i.e., release permits.
* Process for gaseous/liquid release approvals, i.e., release permits.
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures. Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits. SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc. E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedure's overall mitigative strategy or purpose.  
Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.
 
SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.
The applicant's knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedure's content is required to correctly answer the written test item, for example:
E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.
The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:
Page 6 of 16


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 7 of 16
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
* Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
* Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
SRO-only knowledge should not be claimed for questions that can be answered solely using "systems knowledge"; e.g.:
SRO-only knowledge should not be claimed for questions that can be answered solely using systems knowledge; e.g.:
* how the system works.
* how the system works.
* system flow path.
* system flow path.
* component locations, etc. SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:
* component locations, etc.
SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:
* the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
* the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
* any AOP entry condition.
* any AOP entry condition.
* plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
* plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
* immediate operator actions of a procedure. Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.        
* immediate operator actions of a procedure.
Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.
Page 7 of 16


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 8 of 16  Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5) (Assessment and selection of procedures)         
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)
(Assessment and selection of procedures)
Can the question be answered solely by knowing systems knowledge, i.e., how the system works,      Yes RO question flowpath, logic, component location?
No Can the question be answered solely by knowing immediate operator actions?                            Yes    RO question No Can the question be answered solely by knowing entry conditions for AOPs or plant parameters          Yes RO question that require direct entry to major EOPs?
No Can the question be answered solely by knowing the purpose, overall sequence of events, or            Yes RO question overall mitigative strategy of a procedure?
No                                                            The question requires that the applicant select an Does the question require one or more of the following?
SOI section.
* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps                                            Yes       SRO-only
* Knowledge of diagnostic steps and decision points in the              question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16


Can the question be answered solely by knowing "systems knowledge", i.e., how the system works, flowpath, logic, component location? RO question YesNoCan the question be answered solely by knowing immediate operator actions? YesCan the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?YesNoDoes the question require one or more of the following?
93.
* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
00:00:00 -       Unit 1 is at 100% power.
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps
                    -   Unit 2 requests that SCCW be taken out of service to permit maintenance on the Unit 2 cooling tower (CT) basin.
* Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
0 1:00:00     -   The crew performs the following step of Section 7.1, Shutdown, of SOI-27.03:
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?NoNoYesQuestion might not be linked to 10 CFR 55.43(b)(5) for SRO-only NoRO question RO question RO question 
04:00:00 -       Section 7.1 of SOI-27.03 is complete.
: 93. 00:00:00 - Unit 1 is at 100% power. - Unit 2 requests that SCCW be taken out of service to permit maintenance on the Unit 2 cooling tower (CT) basin. 0 1:00:00 - The crew performs the following step of Section 7.1, Shutdown, of SOI-27.03: 04:00:00 - Section 7.1 of SOI-27.03 is complete. 1 1 :00:00 - The Unit 2 CT basin is isolated and drained. - Due to evaporative losses, the Unit 1 CT basin REQUIRES makeup.
1 1 :00:00 -   The Unit 2 CT basin is isolated and drained.
                    -   Due to evaporative losses, the Unit 1 CT basin REQUIRES makeup.
Which ONE of the following describes which document for which the MINIMUM Flow is a basis AND the method for providing makeup to the Unit 1 CT?
Which ONE of the following describes which document for which the MINIMUM Flow is a basis AND the method for providing makeup to the Unit 1 CT?
The source of the MINIMUM flow requirement described in step [1] shown above is ____(1)____. At 1 1 :00:0 1, the crew can provide makeup to the Unit 1 CT using ____(2)_____. A. (1) the bases for T/S LCO 3.7.9 (2) ONLY the RCW system using section 8.3.1 of 0-SOI-24.01 B. (1) the NPDES permit for WBNP (2) ONLY the RCW system using section 8.3.1 of 0-SOI-24.01 C. (1) the bases for T/S LCO 3.7.9 (2) EITHER the SCCW supply using section 8.7 of SOI-27.03 OR the RCW system using section 8.3.1 of 0-SOI-24.01 D. (1) the NPDES permit for WBNP (2) EITHER the SCCW supply using section 8.7 of SOI-27.03 OR the RCW system using section 8.3.1 of 0-SOI-24.01   NOTE: SOI-27.03, Supplemental Condenser Circulating Water System  0-SOI-24.01, Raw Cooling Water System SOI-27.03, Section 8.7, Cooling Tower Basin Makeup with SCCW Shutdown 0-SOI-24.01 section 8.3.1, Unit 1 Bypass Strainer Operation-RCW Adjustment  T/S LCO 3.7.9, Ultimate Heat Sink CORRECT ANSWER:B DISTRACTOR ANALYSIS:  A. Incorrect: Section 8.7 of SOI-27.03 uses the SCCW system to provide makeup to the cooling tower basins. It utilizes a valve named 0-FCV-27-112.
The source of the MINIMUM flow requirement described in step [1] shown above is ____(1)____.
This valve admits water to the U2 cooling tower basin. Because the U2 cooling tower basin is drained for maintenance, it is correct that 0-FCV-27-112 could not be used to compensate for cooling tower evaporative loses; it is not correct and yet plausible to believe that the basis for T/S LCO 3.7.9 is the source for the step [1] cited from Section 7.1, "Shutdown" of SOI-27.03. B. Correct: T/S LCO 3.7.9 stipulates that, "The UHS shall be OPERABLE."  The basis for this LCO indicates:  "The UHS is required to be OPERABLE and is considered OPERABLE if it contains water at or below the maximum temperature that would allow the ERCW System to operate for at least 30 days following the design basis LOCA-To meet this condition, the UHS temperature should not exceed 85&deg;F."  The basis for this T/S does not mention any required river flow.
At 1 1 :00:0 1, the crew can provide makeup to the Unit 1 CT using ____(2)_____.
TVA must regulate discharges to the waters of the United States in accordance with the National Pollutant Discharge Elimination System (NPDES). As seen on page 8 of the TVA-Watts Bar Nuclear Plant NPDES Permit TN0020168, "All changes to the flow rate of the SCCW discharge (Outfall 113) shall be done during periods when flow in the receiving waters is at a minimum of 3,500 cubic feet per second-.The thermal mixing zone area has been modified and redefined for this permit-The discharge from Outfall 113 shall be limited and monitored by the permittee."  Therefore, the source of the step [1] cited from Section 7.1, "Shutdown" of SOI-27.03 is the site's NPDES permit. It is correct that ONLY the RCW system using section 8.3.1 of 0-SOI-24.01 can be used for cooling tower makeup. C. Incorrect: While it is correct that the NPDES is the source for the step [1] cited from Section 7.1, "Shutdown" of SOI-27.03, it is not correct and yet plausible that 0-FCV-27-112 could be used to compensate for cooling tower evaporative loses. D. Incorrect:  It is correct that the NPDES permit is the source for the step [1] cited from Section 7.1, "Shutdown" of SOI-27.03.
NOTE:            SOI-27.03, Supplemental Condenser Circulating Water System 0-SOI-24.01, Raw Cooling Water System SOI-27.03, Section 8.7, Cooling Tower Basin Makeup with SCCW Shutdown 0-SOI-24.01 section 8.3.1, Unit 1 Bypass Strainer Operation-RCW Adjustment T/S LCO 3.7.9, Ultimate Heat Sink A.         (1)   the bases for T/S LCO 3.7.9 (2)   ONLY the RCW system using section 8.3.1 of 0-SOI-24.01 B.         (1)   the NPDES permit for WBNP (2)   ONLY the RCW system using section 8.3.1 of 0-SOI-24.01 C.         (1)   the bases for T/S LCO 3.7.9 (2)   EITHER the SCCW supply using section 8.7 of SOI-27.03 OR the RCW system using section 8.3.1 of 0-SOI-24.01 D.         (1)   the NPDES permit for WBNP (2)   EITHER the SCCW supply using section 8.7 of SOI-27.03 OR the RCW system using section 8.3.1 of 0-SOI-24.01
As seen on print 1-47W-831-1, 0-FCV-27-112, admits water from upstream of the Watts Bar dam to the Unit 2 cooling tower basin. This water flows toward the Unit 2 CCW inlet via the Unit 2 flume. During makeup via this means the Unit 1 and Unit 2 flumes are cross connected. Therefore, the makeup admitted via 0-FCV-27-112 is provided to both Unit 2 and Unit 1. Therefore, it is incorrect that this mode of cooling tower makeup would be utilized (because the Unit 2 cooling tower is drained for maintenance). On the same print one may observe the Unit 1 RCW discharge (30" line) to the Unit 1 cooling tower flume. This line not only provides the discharge of the RCW system but also any bypass strainer flow. The later was the cooling tower makeup afforded by the original plant design (e.g. before the installation of SCCW). The operating crew could divert some of the RCW supply directly to the cooling tower flumes to provide makeup to the cooling tower basins and thus maintain cooling tower levels. If one believed that 0-FCV-27-112 supplied the Unit 1 cooling tower basin, then this answer would be entirely correct.
Question Number: 93  Tier:  2 Group:  2 K/A: 075 Circulating Water System 2.1 Conduct of Operations 2.1.20 Ability to interpret and execute procedure steps. Importance Rating: 4.6  4.6


10 CFR Part 55: (CFR: 41.10 / 43.5 / 45.12) 10CFR55.43.b: 10 CFR 55.43(b)(1)(2) K/A Match: K/A is matched because the applicant must tell the meaning of (i.e. interpret) a component of step [1] of Section 7.1, "Shutdown" of SOI-27.03. Additionally, the applicant must select the correct SOI section in order to provide makeup water to the cooling tower basin. Therefore, the applicant must interpret what the steps in the sections will perform. Technical  
CORRECT ANSWER:                                                              B DISTRACTOR ANALYSIS:
A. Incorrect: Section 8.7 of SOI-27.03 uses the SCCW system to provide makeup to the cooling tower basins. It utilizes a valve named 0-FCV-27-112.
This valve admits water to the U2 cooling tower basin. Because the U2 cooling tower basin is drained for maintenance, it is correct that 0-FCV-27-112 could not be used to compensate for cooling tower evaporative loses; it is not correct and yet plausible to believe that the basis for T/S LCO 3.7.9 is the source for the step [1] cited from Section 7.1, Shutdown of SOI-27.03.
B. Correct: T/S LCO 3.7.9 stipulates that, The UHS shall be OPERABLE. The basis for this LCO indicates: The UHS is required to be OPERABLE and is considered OPERABLE if it contains water at or below the maximum temperature that would allow the ERCW System to operate for at least 30 days following the design basis LOCATo meet this condition, the UHS temperature should not exceed 85&deg;F. The basis for this T/S does not mention any required river flow.
TVA must regulate discharges to the waters of the United States in accordance with the National Pollutant Discharge Elimination System (NPDES). As seen on page 8 of the TVA-Watts Bar Nuclear Plant NPDES Permit TN0020168, All changes to the flow rate of the SCCW discharge (Outfall 113) shall be done during periods when flow in the receiving waters is at a minimum of 3,500 cubic feet per second.The thermal mixing zone area has been modified and redefined for this permitThe discharge from Outfall 113 shall be limited and monitored by the permittee. Therefore, the source of the step [1] cited from Section 7.1, Shutdown of SOI-27.03 is the sites NPDES permit.
It is correct that ONLY the RCW system using section 8.3.1 of 0-SOI-24.01 can be used for cooling tower makeup.
C. Incorrect: While it is correct that the NPDES is the source for the step [1] cited from Section 7.1, Shutdown of SOI-27.03, it is not correct and yet plausible that 0-FCV-27-112 could be used to compensate for cooling tower evaporative loses.
D. Incorrect: It is correct that the NPDES permit is the source for the step [1] cited from Section 7.1, Shutdown of SOI-27.03.
As seen on print 1-47W-831-1, 0-FCV-27-112, admits water from upstream of the Watts Bar dam to the Unit 2 cooling tower basin. This water flows toward the Unit 2 CCW inlet via the Unit 2 flume. During makeup via this means the Unit 1 and Unit 2 flumes are cross connected. Therefore, the makeup admitted via 0-FCV-27-112 is provided to both Unit 2 and Unit 1. Therefore, it is incorrect that this mode of cooling tower makeup would be utilized (because the Unit 2 cooling tower is drained for maintenance). On the same print one may
 
observe the Unit 1 RCW discharge (30 line) to the Unit 1 cooling tower flume. This line not only provides the discharge of the RCW system but also any bypass strainer flow. The later was the cooling tower makeup afforded by the original plant design (e.g. before the installation of SCCW). The operating crew could divert some of the RCW supply directly to the cooling tower flumes to provide makeup to the cooling tower basins and thus maintain cooling tower levels. If one believed that 0-FCV-27-112 supplied the Unit 1 cooling tower basin, then this answer would be entirely correct.
 
Question Number:        93 Tier:    2  Group:        2 K/A:    075 Circulating Water System 2.1 Conduct of Operations 2.1.20 Ability to interpret and execute procedure steps.
Importance Rating:      4.6 4.6 10 CFR Part 55:       (CFR: 41.10 / 43.5 / 45.12) 10CFR55.43.b:         10 CFR 55.43(b)(1)(2)
K/A Match:   K/A is matched because the applicant must tell the meaning of (i.e.
interpret) a component of step [1] of Section 7.1, Shutdown of SOI-27.03. Additionally, the applicant must select the correct SOI section in order to provide makeup water to the cooling tower basin. Therefore, the applicant must interpret what the steps in the sections will perform.
Technical  


==Reference:==
==Reference:==
0-SOI-24.01, Raw Cooling Water System SOI-27.03, Supplemental Condenser Circulating Water System T/S LCO 3.7.9 T/S LCO 3.7.9 Basis 1-47W831-1 NPDES Permit for WBN Proposed references to be provided: None  Learning Objective: 3-OT-SYS027A 10. Given plant conditions, IDENTIFY the applicable Condenser Circulating Water System Precautions and Limitations related to the following: SOI-27.01 Condenser Circulating Water System SOI-27.03 Supplemental CCW System   Cognitive Level:     Higher X Lower     Question Source:     New X Modified Bank   Bank     Question History: New question for the 2015-301 NRC SRO Exam Comments: See the marked up Clarification Guidance for SRO-only Questions.
0-SOI-24.01, Raw Cooling Water System SOI-27.03, Supplemental Condenser Circulating Water System T/S LCO 3.7.9 T/S LCO 3.7.9 Basis 1-47W831-1 NPDES Permit for WBN Proposed references to       None be provided:
UHS 3.7.9         Watts Bar-Unit 1 3.7-21      3.7  PLANT SYSTEMS 3.7.9 Ultimate Heat Sink (UHS)   LCO 3.7.9 The UHS shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. UHS inoperable. A.1 Be in MODE 3. AND A.2 Be in MODE 5. 6 hours    36 hours     SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.9.1 Verify average water temperature of UHS is 85F. 24 hours UHS B 3.7.9  BASES    (continued)      Watts Bar-Unit 1 B 3.7-49      APPLICABLE  (Ref. 2), which requires a 30 day supply of cooling water in the UHS. SAFETY ANALYSES (continued)  The UHS satisfies Criterion 3 of the NRC Policy Statement.
Learning Objective:         3-OT-SYS027A
: 10. Given plant conditions, IDENTIFY the applicable Condenser Circulating Water System Precautions and Limitations related to the following:
SOI-27.01 Condenser Circulating Water System SOI-27.03 Supplemental CCW System Cognitive Level:
Higher               X Lower Question Source:
New                   X Modified Bank Bank Question History:           New question for the 2015-301 NRC SRO Exam Comments:                   See the marked up Clarification Guidance for SRO-only Questions.
 
UHS 3.7.9 Nothing above the 3.7 PLANT SYSTEMS                                                             line for this T/S leads one to the 3.7.9 Ultimate Heat Sink (UHS) correct answer.
LCO 3.7.9               The UHS shall be OPERABLE.
APPLICABILITY:         MODES 1, 2, 3, and 4.
ACTIONS CONDITION                           REQUIRED ACTION           COMPLETION TIME A.       UHS inoperable.             A.1     Be in MODE 3.             6 hours AND A.2     Be in MODE 5.             36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE                                 FREQUENCY SR 3.7.9.1             Verify average water temperature of UHS is 85 F. 24 hours Watts Bar-Unit 1                               3.7-21


LCO   The UHS is required to be OPERABLE and is considered OPERABLE if it contains water at or below the maximum temperature that would allow the ERCW System to operate for at least 30 days following the design basis LOCA without the loss of net positive suction head (NPSH), and without exceeding the maximum design temperature of the equipment served by the ERCW System. To meet this condition, the UHS temperature should not exceed 85F.
UHS B 3.7.9 BASES APPLICABLE      (Ref. 2), which requires a 30 day supply of cooling water in the UHS.
APPLICABILITY In MODES 1, 2, 3, and 4, the UHS is required to support the OPERABILITY of the equipment serviced by the UHS and required to be OPERABLE in these MODES.
SAFETY ANALYSES (continued)    The UHS satisfies Criterion 3 of the NRC Policy Statement.
In MODE 5 or 6, the OPERABILITY requirements of the UHS are determined by the systems it supports. ACTIONS A.1 If the UHS is inoperable, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours and in MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.7.9.1 REQUIREMENTS This SR verifies that the ERCW System is available to cool the CCS to at least its maximum design temperature with the maximum accident or normal design heat loads for 30 days following a Design Basis Accident. The 24 hour Frequency
LCO             The UHS is required to be OPERABLE and is considered OPERABLE if it contains water at or below the maximum temperature that would allow the ERCW System to operate for at least 30 days following the design basis LOCA without the loss of net positive suction head (NPSH), and without exceeding the maximum design temperature of the equipment served by the ERCW System.
To meet this condition, the UHS temperature should not exceed 85 F.
APPLICABILITY   In MODES 1, 2, 3, and 4, the UHS is required to support the OPERABILITY of the equipment serviced by the UHS and required to be OPERABLE in these MODES.
In MODE 5 or 6, the OPERABILITY requirements of the UHS are determined by the systems it supports.
ACTIONS         A.1 If the UHS is inoperable, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours and in MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE     SR 3.7.9.1 REQUIREMENTS This SR verifies that the ERCW System is available to cool the CCS to at least its maximum design temperature with the maximum accident or normal design heat loads for 30 days following a Design Basis Accident. The 24 hour Frequency (continued)
Watts Bar-Unit 1                          B 3.7-49


TVA-Watts-BarNuclearPlantNPDESPermitTN0020168Page8of27DischargesareauthorizedforOutfall101onlyduringperiodswhenflowinthereceivingstreamisataminimumof3,500cubicfeetpersecond.AllchangestotheflowrateoftheSCCWdischarge(Outfall113)shallbedoneduringperiodswhenflowinthereceivingwatersisataminimumof3,500Cubicfeetpersecond.Thisincludesperiodsofstart-up,shutdownaswellasothersimilarabruptflowratechangesoftheSCCW.WhenthermallyloadedeffluentisdischargedthroughOutfall102,allreasonableeffortsshallbemadetokeepflowtoaminimumof3500cubicfeetpersecondinthereceivingwaters.Ifsuchflowisabsent,thepermitteeshallverifyprotectionofwaterqualitybytakinginstreamtemperaturemeasurements.Compliancewithflowrequirementsfor3,500cfsflowinstreamforOutfalls101,102and113dischargesshallbecertifiedmonthlywiththesubmissionofDischargeMonitoringReportssubmittedtotheDivisionfortheseoutfalls.Recordsconcerningtheinstreamflowshallbemaintainedandavailableuponrequest.Thethermalmixingzoneareahasbeenmodifiedandredefinedforthispermit;seediagramatAppendix5H.ThedischargefromOutfall113shallbelimitedandmonitoredbythepermitteeasspecifiedbelow:*Inrecognitionofthedynamicbehaviorofthethermaleffluentintheriver,monitoringshallberequiredforanactivemixingzoneandapassivemiXingzoneasdescribedinthepermitrationale.Thepassivemixingzoneincludesthefollowingdimensions:(1)amaximumwidthoffrombanktobankintheriver,and(2)amaximumlengthof1000feetdownstreamoftheoutfall.Ithasbeendocumentedthatthereisazoneof(coolwater)refugeinthebottomlayertoallowforfishandotherspeciestopassbelowthethermalplume.Compliancewiththerequirementsbelowwillbeestablishedfortheactivemixingzoneatamaximumlengthof2000feetdownstreamoftheoutfall.*ComplianceforthepassivemiXingzoneshallbebytwoinstreamtemperaturesurveys,oneconductedduringwinterambientconditionsandoneduringsummerambientconditions.ThesurveysshallbeperformedwhiletheSCCWsystemisthermallyloadedwithlowriver.flowconditionsandshallincludetemperatureprofilesatasufficientnumberoflocationsacrossthedownstreamedgeofthepassivemixingzonetolocatetheeffluentplume.Themeasurementsshallbecomparedwiththeresultsfromthethermalplumemodelandshallbesummarizedinareporttothedivisionsemiannually.*CompliancewithTEMPERATURE,EdgeofMixingzone;TEMPERATURE,RiseUpstreamtoDownstream;andTEMPERATURE,RateofChangeshallbeapplicableattheedgeoftheactivemixingzone.*DailymaximumtemperaturesfortheTEMPERATURE,effluent;TEMPERATURE,EdgeofMixingzone;TEMPERATURE,RiseUpstreamtoDownstream;andTEMPERATURE,RateofChangeshallbedeterminedfrom1-houraveragevalues.Theaveragevaluesshallbecalculatedevery15minutesusingthecurrentandpreviousfour15-minutevalues,thuscreatingarollingaverage.*Asdemonstratedbymonitoringattheedgeoftheactivemixingzone,themaximumtemperatureshallnotexceed30.5&deg;C(exceptasaresultofnaturalcauses),themaximumchangeintemperaturerelativetotheupstreamcontrolpointshallnotexceed3&deg;C(exceptasaresultofnaturalcauses),andthemaximumtemperaturerateofchangeshallnotexceed2&deg;Cperhour(exceptasaresultofnaturalcauses).**
TVA-Watts-Bar Nuclear Plant NPDES Permit TN0020168 Page 8 of 27  *
* Discharges are authorized for Outfall 101 only during periods when flow in the receiving stream is at a minimum of 3,500 cubic feet per second. All changes to the flow rate of the SCCW discharge (Outfall 113) shall be done during periods when flow in the receiving waters is at a minimum of 3,500 Cubic feet per second. This includes periods of start-up, shutdown as well as other similar abrupt flow rate changes of the SCCW. When thermally loaded effluent is discharged through Outfall 102, all reasonable efforts shall be made to keep flow to a minimum of 3500 cubic feet per second in the receiving waters. If such flow is absent, the permittee shall verify protection of water quality by taking instream temperature measurements. Compliance with flow requirements for 3,500 cfs flow instream for Outfalls 101, 102 and 113 discharges shall be certified monthly with the submission of Discharge Monitoring Reports submitted to the Division for these outfalls. Records concerning the instream flow shall be maintained and available upon request.
The thermal mixing zone area has been modified and redefined for this permit; see diagram at Appendix 5H. The discharge from Outfall 113 shall be limited and monitored by the permittee as specified below:
* In recognition of the dynamic behavior of the thermal effluent in the river, monitoring shall be required for an active mixing zone and a passive miXing zone as described in the permit rationale. The passive mixing zone includes the following dimensions:
(1) a maximum width of from bank to bank in the river, and (2) a maximum length of 1000 feet downstream of the outfall. It has been documented that there is a zone of (cool water) refuge in the bottom layer to allow for fish and other species to pass below the thermal plume. Compliance with the requirements below will be established for the active mixing zone at a maximum length of 2000 feet downstream of the outfall.
* Compliance for the passive miXing zone shall be by two instream temperature surveys, one conducted during winter ambient conditions and one during summer ambient conditions. The surveys shall be performed while the SCCW system is thermally loaded with low river .flow conditions and shall include temperature profiles at a sufficient number of locations across the downstream edge of the passive mixing zone to locate the effluent plume. The measurements shall be compared with the results from the thermal plume model and shall be summarized in a report to the division semiannually.
* Compliance with TEMPERATURE, Edge of Mixing zone; TEMPERATURE, Rise Upstream to Downstream; and TEMPERATURE, Rate of Change shall be applicable at the edge of the active mixing zone.
* Daily maximum temperatures for the TEMPERATURE, effluent; TEMPERATURE, Edge of Mixing zone; TEMPERATURE, Rise Upstream to Downstream; and TEMPERATURE, Rate of Change shall be determined from 1-hour average values.
The average values shall be calculated every 15 minutes using the current and previous four 15-minute values, thus creating a rolling average.
* As demonstrated by monitoring at the edge of the active mixing zone, the maximum temperature shall not exceed 30.5&deg;C (except as a result of natural causes), the maximum change in temperature relative to the upstream control point shall not exceed 3&deg;C (except as a result of natural causes), and the maximum temperature rate of change shall not exceed 2&deg;C per hour (except as a result of natural causes).


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 3 of 16  II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]: A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)] Some examples of SRO exam items for this topic include:
asin Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]:
A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)]
Some examples of SRO exam items for this topic include:
* Reporting requirements when the maximum licensed thermal power output is exceeded.
* Reporting requirements when the maximum licensed thermal power output is exceeded.
* Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems, fire doors, etc.
* Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems, fire doors, etc.
* The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
* The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
* National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
* National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
* Processes for TS and FSAR changes. Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic. B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)] Some examples of SRO exam items for this topic include:
* Processes for TS and FSAR changes.
* Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic.
B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]
Some examples of SRO exam items for this topic include:
The applicant must understand what
* Application of Required Actions (Section 3) and Surveillance constitutes an            Requirements (SR) (Section 4) in accordance with rules of application OPERABLE UHS.            requirements (Section 1).
* Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
* Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
* Knowledge of TS bases that are required to analyze TS required actions and terminology.
* Knowledge of TS bases that are required to analyze TS required actions and terminology.
* Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).
* Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of  1 hour action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.  
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of  1 hour action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. ROs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Page 3 of 16


SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. RO's are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 4 of 16 ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)         Above this line RO knowledge Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 5 of 16 Figure 1:  Screening for SRO-only linked to 10 CFR 55.43(b)(2) (Tech Specs)       
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)
Nothing "above the line" for the UHS T/S leads the applicant to the RO                                                      correct answer.
knowledge Above this line Page 4 of 16


Can question be answered solely by knowing  1 hour TS/TRM Action? RO question YesNoCan question be answered solely by knowing the LCO/TRM information listed "above-the-line?" YesRO question NoCan question be answered solely by knowing the TS Safety Limits? YesRO question NoDoes the question involve one or more of the following for TS, TRM, or ODCM?
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)
(Tech Specs)
Can question be answered solely by knowing  1       Yes hour TS/TRM Action?                                           RO question No Can question be answered solely by knowing the       Yes LCO/TRM information listed above-the-line?                   RO question No Can question be answered solely by knowing the       Yes TS Safety Limits?                                             RO question No Does the question involve one or more of the following for TS, TRM, or ODCM?
* Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
* Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)
* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)                             Yes      SRO-only
* Knowledge of TS bases that is required to analyze TS required actions and terminology SRO-only question YesNoQuestion might not be linked to 10 CFR 55.43(b)(2) for SRO-only
* Knowledge of TS bases that is required to analyze TS                   question required actions and terminology No Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only Page 5 of 16
: 94. Given the following conditions:  -  Unit 1 is stable in MODE 2.
-  Chemistry reports that DOSE EQUIVALENT I-131 (DEI) is 1.0 &#xb5;Ci/gm. Excerpt from T/S LCO 3.4.16 Which ONE of the following describes the application of the Technical Specifications for the conditions below?


94.
Given the following conditions:
    - Unit 1 is stable in MODE 2.
    - Chemistry reports that DOSE EQUIVALENT I-131 (DEI) is 1.0 &#xb5;Ci/gm.
Excerpt from T/S LCO 3.4.16 Which ONE of the following describes the application of the Technical Specifications for the conditions below?
In accordance with T/S LCO required action 3.4.16 A.1, DEI must be verified < ____(1)____ &#xb5;Ci/gm.
In accordance with T/S LCO required action 3.4.16 A.1, DEI must be verified < ____(1)____ &#xb5;Ci/gm.
The NOTE "LCO 3.0.4.c is applicable" indicates that when an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made _____(2)_____.   (1) (2) A. 14 when an allowance is stated in the individual value, parameter or other Specification B. 14 after performance of a risk assessment addressing inoperable systems and components, and establishment of risk management actions C. 21 when an allowance is stated in the individual value, parameter or other Specification D. 21 after performance of a risk assessment addressing inoperable systems and components, and establishment of risk management actions CORRECT ANSWER:A DISTRACTOR ANALYSIS:   A. Correct: A verification that DEI is <14 &#xb5;Ci/gm must be made. Additionally, as discussed a risk assessment is not required to support the mode 1 change.
The NOTE LCO 3.0.4.c is applicable indicates that when an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made _____(2)_____.
As seen in the required action A.1 for T/S LCO 3.4.16, a verification must be made that DEI is <14 &#xb5;Ci/gm. This limit is contained throughout the basis for T/S LCO 3.4.16. LCO 3.0.4 states: "When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made: a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time;  
(1)                                             (2)
: b. After performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate; exceptions to this Specification are stated in the individual Specifications, or  
A.           14       when an allowance is stated in the individual value, parameter or other Specification B.           14       after performance of a risk assessment addressing inoperable systems and components, and establishment of risk management actions C.           21       when an allowance is stated in the individual value, parameter or other Specification D.           21       after performance of a risk assessment addressing inoperable systems and components, and establishment of risk management actions
: c. When an allowance is stated in the individual value, parameter, or other Specification."
 
One may see that the required actions for condition A of T/S LCO 3.4.16 are modified by the NOTE which states: "LCO 3.0.4.c is applicable."  Because of this, entry into mode 1 is not impeded by the fact that condition A of T/S LCO is not met.
CORRECT ANSWER:                                                                 A DISTRACTOR ANALYSIS:
B. Incorrect: As seen in the required action A.1 for T/S LCO 3.4.16, a verification must be made that DEI is <14 &#xb5;Ci/gm. This limit is contained throughout the basis for T/S LCO 3.4.16. As such, the first half of this distractor is correct. It is plausible to believe that a risk assessment would be required because this would be the case if LCO 3.0.4 b were invoked to support the mode change. C. Incorrect: It is not correct that a verification must be made that DEI is <21 &#xb5;Ci/gm. It is plausible to believe that this is the case because prior to amendment 91, T/S LCO 3.4.16 (and its basis) utilized this value. This amendment was placed into effect during the SRO applicant's time in initial license training.
A. Correct: A verification that DEI is <14 &#xb5;Ci/gm must be made. Additionally, as discussed a risk assessment is not required to support the mode 1 change.
It is Correct: however, that a risk assessment is not required prior to an entry into mode 1. D. Incorrect: It is not correct that a verification must be made that DEI is <21 &#xb5;Ci/gm. It is plausible to believe that this is the case because prior to amendment 91, T/S LCO 3.4.16 (and its basis) utilized this value. This amendment was placed into effect during the SRO applicants time in initial license training.
As seen in the required action A.1 for T/S LCO 3.4.16, a verification must be made that DEI is <14 &#xb5;Ci/gm. This limit is contained throughout the basis for T/S LCO 3.4.16.
LCO 3.0.4 states: When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made:
: a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time;
: b. After performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate; exceptions to this Specification are stated in the individual Specifications, or
: c. When an allowance is stated in the individual value, parameter, or other Specification.
One may see that the required actions for condition A of T/S LCO 3.4.16 are modified by the NOTE which states: LCO 3.0.4.c is applicable. Because of this, entry into mode 1 is not impeded by the fact that condition A of T/S LCO is not met.
B. Incorrect: As seen in the required action A.1 for T/S LCO 3.4.16, a verification must be made that DEI is <14 &#xb5;Ci/gm. This limit is contained throughout the basis for T/S LCO 3.4.16. As such, the first half of this distractor is correct. It is plausible to believe that a risk assessment would be required because this would be the case if LCO 3.0.4 b were invoked to support the mode change.
C. Incorrect: It is not correct that a verification must be made that DEI is <21
              &#xb5;Ci/gm. It is plausible to believe that this is the case because prior to amendment 91, T/S LCO 3.4.16 (and its basis) utilized this value. This amendment was placed into effect during the SRO applicants time in initial license training.
It is Correct: however, that a risk assessment is not required prior to an entry into mode 1.
D. Incorrect: It is not correct that a verification must be made that DEI is <21
              &#xb5;Ci/gm. It is plausible to believe that this is the case because prior to amendment 91, T/S LCO 3.4.16 (and its basis) utilized this value. This amendment was placed into effect during the SRO applicants time in initial license training.
Additionally, as previously discussed, it is incorrect and yet plausible that a risk assessment be required to enter mode 1.
Additionally, as previously discussed, it is incorrect and yet plausible that a risk assessment be required to enter mode 1.
Question Number: 94  Tier:  3  Group:
K/A: 2.1 Conduct of Operations 2.1.34 Knowledge of primary and secondary plant chemistry limits. Importance Rating: 2.7  3.5 10 CFR Part 55: (CFR: 41.10 / 43.5 / 45.12)


10CFR55.43.b: 10 CFR 55.43(b)(2) K/A Match: K/A is matched because the applicant is required to possess the knowledge of one of the limits germane to RCS specific activity.
Question Number:      94 Tier:    3    Group:
K/A:    2.1 Conduct of Operations 2.1.34 Knowledge of primary and secondary plant chemistry limits.
Importance Rating:    2.7 3.5 10 CFR Part 55:      (CFR: 41.10 / 43.5 / 45.12) 10CFR55.43.b:       10 CFR 55.43(b)(2)
K/A Match:   K/A is matched because the applicant is required to possess the knowledge of one of the limits germane to RCS specific activity.
Technical  
Technical  


==Reference:==
==Reference:==
T/S LCO 3.4.16 (current and a historical copy)
T/S LCO 3.4.16 (current and a historical copy)
T/S LCO 3.4.16 basis T/S LCO 3.0.4 Proposed references to be provided: None  Learning Objective: 3-OT-TS-0304 1. Describe the LCO, Applicability and Bases for the LCO. Cognitive Level:     Higher   Lower X   Question Source:     New X Modified Bank   Bank     Question History: New question for the 2015-301 NRC SRO Exam Comments:
T/S LCO 3.4.16 basis T/S LCO 3.0.4 Proposed references to   None be provided:
LCO Applicability 3.0      (continued)    Watts Bar-Unit 1 3.0-1 Amendment 55    3.0  LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY  LCO  3.0.1  LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2. LCO  3.0.2  Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6. If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required unless otherwise stated. LCO  3.0.3  When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within 1 hour to place the unit, as applicable, in:  a. MODE 3 within 7 hours;  b. MODE 4 within 13 hours; and  c. MODE 5 within 37 hours. Exceptions to this Specification are stated in the individual Specifications. Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by LCO 3.0.3 is not required. LCO 3.0.3 is only applicable in MODES 1, 2, 3, and 4. LCO  3.0.4  When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made:      a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time; LCO Applicability 3.0      (continued)    Watts Bar-Unit 1 3.0-2 Amendment 55    3.0  LCO APPLICABILITY  LCO  3.0.4  b. After performance of a risk assessment addressing inoperable  (continued)    systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate; exceptions to this Specification are stated in the individual Specifications, or  c. When an allowance is stated in the individual value, parameter, or other Specification. This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. LCO  3.0.5  Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY. LCO  3.0.6  When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, additional evaluations and limitations may be required in accordance with Specification 5.7.2.18, "Safety Function Determination Program (SFDP)."  If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.
Learning Objective:       3-OT-TS-0304
RCS Specific Activity3.4.16        Watts Bar-Unit 1 3.4-39 Amendment 41, 55, 91 3.4  REACTOR COOLANT SYSTEM (RCS) 3.4.16  RCS Specific Activity LCO  3.4.16  The specific activity of the reactor coolant shall be within limits.
: 1. Describe the LCO, Applicability and Bases for the LCO.
APPLICABILITY: MODES 1 and 2,    MODE 3 with RCS average temperature (Tavg)  500F. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. DOSE EQUIVALENT I-131 > 0.265 Ci/gm. 
Cognitive Level:
-----------------------NOTE-------------------- LCO 3.0.4.c is applicable. ---------------------------------------------------
Higher Lower               X Question Source:
A.1 Verify DOSE EQUIVALENT I-131  14 Ci/gm  AND  A.2 Restore DOSE EQUIVALENT I-131 to within limit.
New               X Modified Bank Bank Question History:         New question for the 2015-301 NRC SRO Exam Comments:
Once per 4 hours
 
48 hours  B. Gross specific activity of the reactor coolant not within limit.
B.1 Perform SR 3.4.16.2. AND  B.2 Be in MODE 3 with Tavg < 500F. 4 hours 6 hours  (continued)
RCS Specific Activity3.4.16        Watts Bar-Unit 1 3.4-40 Amendment 41, 91 ACTIONS  (continued) CONDITION REQUIRED ACTION COMPLETION TIME  C. Required Action and associated Completion Time of Condition A not met. OR  DOSE EQUIVALENT I-131          > 14 Ci/gm. C.1 Be in MODE 3 with Tavg < 500F.
6 hours 
 
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR  3.4.16.1  Verify reactor coolant gross specific    activity < 100/E Ci/gm. 7 days SR  3.4.16.2  --------------------------------NOTE-----------------------------------  Only required to be performed in MODE 1.    ---------------------------------------------------------------------------    Verify reactor coolant DOSE EQUIVALENT I-131 specific activity  0.265 Ci/gm.
14 days  AND  Between 2 and 6 hours after a THERMAL POWER change of  15% RTP within a 1 hour period  (continued)
RCS Specific Activity 3.4.16          Watts Bar-Unit 1 3.4-41      SURVEILLANCE REQUIREMENTS  (continued) SURVEILLANCE FREQUENCY  SR  3.4.16.3  ----------------------------------NOTE---------------------------------    Required to be performed within 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for  48 hours.    ---------------------------------------------------------------------------    Determine  from a sample taken in MODE 1 after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for  48 hours.            184 days RCS Specific Activity3.4.16 Watts Bar-Unit 1 3.4-39 Amendment 41, 55 3.4  REACTOR COOLANT SYSTEM (RCS)
 
3.4.16  RCS Specific Activity
 
LCO  3.4.16  The specific activity of the reactor coolant shall be within limits. 
 
APPLICABILITY: MODES 1 and 2,    MODE 3 with RCS average temperature (Tavg)  500F.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. DOSE EQUIVALENT I-131 > 0.265 Ci/gm. 
---------------------NOTE-------------------- LCO 3.0.4.c is applicable. -------------------------------------------------
A.1 Verify DOSE EQUIVALENT I-131  21 Ci/gm  AND A.2 Restore DOSE EQUIVALENT I-131 to within limit. 
 
Once per 4 hours
 
48 hours B. Gross specific activity of the reactor coolant not within limit.
B.1 Perform SR 3.4.16.2.


AND  B.2 Be in MODE 3 with Tavg < 500F. 4 hours
LCO Applicability 3.0 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1          LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2.
LCO 3.0.2          Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6.
If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required unless otherwise stated.
LCO 3.0.3          When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within 1 hour to place the unit, as applicable, in:
: a.      MODE 3 within 7 hours;
: b.      MODE 4 within 13 hours; and
: c.      MODE 5 within 37 hours.
Exceptions to this Specification are stated in the individual Specifications.
Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by LCO 3.0.3 is not required.
LCO 3.0.3 is only applicable in MODES 1, 2, 3, and 4.
LCO 3.0.4          When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made:
: a.      When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time; (continued)
Watts Bar-Unit 1                                3.0-1                                    Amendment 55


6 hours  (continued)
LCO Applicability 3.0 Notice that a,b and 3.0 LCO APPLICABILITY                                                                        c or joined by OR.
LCO 3.0.4                    b.      After performance of a risk assessment addressing inoperable (continued)                          systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate; exceptions to this Specification are stated in the individual Specifications, or
: c.      When an allowance is stated in the individual value, parameter, or other Specification.
This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.
LCO 3.0.5          Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY.
LCO 3.0.6           When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, additional evaluations and limitations may be required in accordance with Specification 5.7.2.18, "Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.
(continued)
Watts Bar-Unit 1                                3.0-2                                      Amendment 55


RCS Specific Activity3.4.16 Watts Bar-Unit 1 3.4-40 Amendment 41 ACTIONS  (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and associated Completion Time of Condition A not met.
RCS Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.16              The specific activity of the reactor coolant shall be within limits.
OR DOSE EQUIVALENT I-131 > 21 Ci/gm. C.1 Be in MODE 3 with Tavg < 500F.
APPLICABILITY:          MODES 1 and 2, MODE 3 with RCS average temperature (Tavg) 500&deg;F.
6 hours   
ACTIONS CONDITION                               REQUIRED ACTION                       COMPLETION TIME A. DOSE EQUIVALENT I-131            -----------------------NOTE--------------------
      > 0.265 Ci/gm.                            LCO 3.0.4.c is applicable.
A.1    Verify DOSE EQUIVALENT I-131               Once per 4 hours 14 Ci/gm AND A.2    Restore DOSE EQUIVALENT                    48 hours I-131 to within limit.
B. Gross specific activity of the      B.1     Perform SR 3.4.16.2.                      4 hours reactor coolant not within limit.
AND B.2      Be in MODE 3 with                         6 hours Tavg < 500&deg;F.
(continued)
Watts Bar-Unit 1                                    3.4-39                                Amendment 41, 55, 91


SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 Verify reactor coolant gross specific   activity < 100/E Ci/gm.
RCS Specific Activity 3.4.16 ACTIONS (continued)
7 days  SR 3.4.16.2 --------------------------------NOTE-----------------------------------
CONDITION                                    REQUIRED ACTION                        COMPLETION TIME C. Required Action and                    C.1          Be in MODE 3 with                    6 hours associated Completion Time                            Tavg < 500&deg;F.
Only required to be performed in MODE 1.   ---------------------------------------------------------------------------  
of Condition A not met.
OR DOSE EQUIVALENT I-131
      > 14 Ci/gm.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                     FREQUENCY SR 3.4.16.1         Verify reactor coolant gross specific                                       7 days activity < 100/E Ci/gm.
SR 3.4.16.2         --------------------------------NOTE-----------------------------------
Only required to be performed in MODE 1.
Verify reactor coolant DOSE EQUIVALENT I-131 specific                      14 days activity  0.265 Ci/gm.
AND Between 2 and 6 hours after a THERMAL POWER change of 15% RTP within a 1 hour period (continued)
Watts Bar-Unit 1                                          3.4-40                                    Amendment 41, 91


Verify reactor coolant DOSE EQUIVALENT I-131 specific activity 0.265 Ci/gm.  
RCS Specific Activity 3.4.16 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE                                                    FREQUENCY SR 3.4.16.3    ----------------------------------NOTE---------------------------------
Required to be performed within 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for 48 hours.
Determine from a sample taken in MODE 1 after a                          184 days minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for 48 hours.
Watts Bar-Unit 1                                      3.4-41


14 days
RCS Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.16              The specific activity of the reactor coolant shall be within limits.
APPLICABILITY:          MODES 1 and 2, MODE 3 with RCS average temperature (Tavg)  500&deg;F.
This is the OLD Limit ACTIONS CONDITION                              REQUIRED ACTION                      COMPLETION TIME A. DOSE EQUIVALENT I-131            ---------------------NOTE--------------------
      > 0.265 &#xb5;Ci/gm.                            LCO 3.0.4.c is applicable.
A.1    Verify DOSE EQUIVALENT I-131              Once per 4 hours 21 &#xb5;Ci/gm AND A.2    Restore DOSE EQUIVALENT                  48 hours I-131 to within limit.
B. Gross specific activity of the    B.1      Perform SR 3.4.16.2.                    4 hours reactor coolant not within limit.
AND B.2      Be in MODE 3 with Tavg < 500&deg;F.          6 hours (continued)
Watts Bar-Unit 1                                    3.4-39                                  Amendment 41, 55


AND Between 2 and 6 hours after a THERMAL POWER change of 15% RTP within a 1 hour period   (continued)
RCS Specific Activity 3.4.16 ACTIONS (continued)
CONDITION                                    REQUIRED ACTION                        COMPLETION TIME C. Required Action and                    C.1          Be in MODE 3 with                    6 hours associated Completion Time                          Tavg < 500&deg;F.
of Condition A not met.
OR DOSE EQUIVALENT I-131 >
21 &#xb5;Ci/gm.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                      FREQUENCY SR 3.4.16.1        Verify reactor coolant gross specific                                        7 days activity < 100/E &#xb5;Ci/gm.
SR 3.4.16.2        --------------------------------NOTE-----------------------------------
Only required to be performed in MODE 1.
Verify reactor coolant DOSE EQUIVALENT I-131 specific                        14 days activity  0.265 &#xb5;Ci/gm.
AND Between 2 and 6 hours after a THERMAL POWER change of 15% RTP within a 1 hour period (continued)
Watts Bar-Unit 1                                          3.4-40                                        Amendment 41


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 3 of 16  II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]: A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)] Some examples of SRO exam items for this topic include:
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]:
A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)]
Some examples of SRO exam items for this topic include:
* Reporting requirements when the maximum licensed thermal power output is exceeded.
* Reporting requirements when the maximum licensed thermal power output is exceeded.
* Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems, fire doors, etc.
* Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems, fire doors, etc.
* The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
* The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
* National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
* National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
* Processes for TS and FSAR changes. Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic. B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)] Some examples of SRO exam items for this topic include:
* Processes for TS and FSAR changes.
Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic.
B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]
Some examples of SRO exam items for this topic include:
* Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
* Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
* Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
* Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
* Knowledge of TS bases that are required to analyze TS required actions and terminology.
* Knowledge of TS bases that are required to analyze TS required actions and terminology.
* Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).
* Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of  1 hour action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.  
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of  1 hour action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. ROs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Page 3 of 16


SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. RO's are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 4 of 16 ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)         Above this line RO knowledge Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 5 of 16 Figure 1:  Screening for SRO-only linked to 10 CFR 55.43(b)(2) (Tech Specs)       
Clarification Guidance for SRO-only Questions Question (in part)                          Rev 1 (03/11/2010) requires "below the ACTIONS from the LCO and associated applicability statements line" knowledge of (standardized TS; see example below) the T/S.
RO knowledge Above this line Page 4 of 16


Can question be answered solely by knowing  1 hour TS/TRM Action? RO question YesNoCan question be answered solely by knowing the LCO/TRM information listed "above-the-line?" YesRO question NoCan question be answered solely by knowing the TS Safety Limits? YesRO question NoDoes the question involve one or more of the following for TS, TRM, or ODCM?
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)
(Tech Specs)
Can question be answered solely by knowing  1       Yes hour TS/TRM Action?                                           RO question No Can question be answered solely by knowing the       Yes LCO/TRM information listed above-the-line?                   RO question No Can question be answered solely by knowing the       Yes TS Safety Limits?                                             RO question No Does the question involve one or more of the following for TS, TRM, or ODCM?
* Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
* Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)
* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)                             Yes      SRO-only
* Knowledge of TS bases that is required to analyze TS required actions and terminology SRO-only question YesNoQuestion might not be linked to 10 CFR 55.43(b)(2) for SRO-only
* Knowledge of TS bases that is required to analyze TS                   question required actions and terminology No Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only Page 5 of 16
: 95. In accordance with Tech Spec LCO 3.0.5, INOPERABLE equipment may be returned to service for the following reasons:  
 
: 1. Demonstrate OPERABILITY of the equipment.  
95.
: 2. Demonstrate OPERABILITY of other Tech Spec required equipment. 3. Troubleshoot equipment to facilitate repair. A. 1 ONLY B. 1 and 2 ONLY C. 1 and 3 ONLY D. 1, 2 and 3 CORRECT ANSWER:B DISTRACTOR ANALYSIS:   A. Incorrect: LCO 3.0.5 states: "Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment."  Therefore, this answer is plausible because it is partly correct (i.e. that it is correct that LCO 3.0.5 may be used to demonstrate the OPERABILITY of the inoperable equipment) but it is not fully correct because the OPERABILITY of other equipment may be tested. . B. Correct: T/S LCO states: "Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. "  C. Incorrect: As mentioned, this distractor is incorrect. It is plausible because troubleshooting is part of the process which restores a failed component to an operable status. One may believe that the process by which a component is repaired and thus returned to an operable status is addressed by this T/S. This is not the case. D. Incorrect: Again, while two of the three items listed in this distractor are correct, troubleshooting is not allowed by T/S LCO 3.0.5.
In accordance with Tech Spec LCO 3.0.5, INOPERABLE equipment may be returned to service for the following reasons:
Question Number: 95  Tier:  3 Group:
: 1. Demonstrate OPERABILITY of the equipment.
K/A: 2.2 Equipment Control 2.2.21 Knowledge of pre- and post-maintenance operability requirements. Importance Rating: 2.9  4.1 10 CFR Part 55: (CFR: 41.10 / 43.2)
: 2. Demonstrate OPERABILITY of other Tech Spec required equipment.
: 3. Troubleshoot equipment to facilitate repair.
A. 1 ONLY B. 1 and 2 ONLY C. 1 and 3 ONLY D. 1, 2 and 3
 
CORRECT ANSWER:                                                               B DISTRACTOR ANALYSIS:
A. Incorrect: LCO 3.0.5 states: Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. Therefore, this answer is plausible because it is partly correct (i.e. that it is correct that LCO 3.0.5 may be used to demonstrate the OPERABILITY of the inoperable equipment) but it is not fully correct because the OPERABILITY of other equipment may be tested.
B. Correct: T/S LCO states:
Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment.
C. Incorrect: As mentioned, this distractor is incorrect. It is plausible because troubleshooting is part of the process which restores a failed component to an operable status. One may believe that the process by which a component is repaired and thus returned to an operable status is addressed by this T/S. This is not the case.
D. Incorrect: Again, while two of the three items listed in this distractor are correct, troubleshooting is not allowed by T/S LCO 3.0.5.


10CFR55.43.b: 10 CFR 55.43(b)(2) K/A Match: K/A is matched because the applicant is required to understand the use of T/S LCO 3.0.5 to an inoperable safety system or component.
Question Number:      95 Tier:    3  Group:
K/A:    2.2 Equipment Control 2.2.21 Knowledge of pre- and post-maintenance operability requirements.
Importance Rating:    2.9 4.1 10 CFR Part 55:      (CFR: 41.10 / 43.2) 10CFR55.43.b:       10 CFR 55.43(b)(2)
K/A Match:   K/A is matched because the applicant is required to understand the use of T/S LCO 3.0.5 to an inoperable safety system or component.
Technical  
Technical  


==Reference:==
==Reference:==
T/S LCO 3.0.5 Proposed references to be provided: None  Learning Objective: 3-OT-TS-0300 5. Given plant conditions where LCOs and/or TRs are not met, determine if equipment may be tested to demonstrate operability. Cognitive Level:     Higher   Lower X   Question Source:     New   Modified Bank   Bank X     Question History: Bank question taken verbatim from the last SQN NRC exam. Comments:
T/S LCO 3.0.5 Proposed references to   None be provided:
LCO Applicability 3.0     (continued)    Watts Bar-Unit 1 3.0-2 Amendment 55    3.0 LCO APPLICABILITY   LCO 3.0.4   b. After performance of a risk assessment addressing inoperable   (continued)   systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate; exceptions to this Specification are stated in the individual Specifications, or c. When an allowance is stated in the individual value, parameter, or other Specification. This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY. LCO 3.0.6 When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, additional evaluations and limitations may be required in accordance with Specification 5.7.2.18, "Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.
Learning Objective:       3-OT-TS-0300
: 5. Given plant conditions where LCOs and/or TRs are not met, determine if equipment may be tested to demonstrate operability.
Cognitive Level:
Higher Lower               X Question Source:
New Modified Bank Bank               X Question History:         Bank question taken verbatim from the last SQN NRC exam.
Comments:
 
LCO Applicability 3.0 3.0 LCO APPLICABILITY LCO 3.0.4                   b.       After performance of a risk assessment addressing inoperable (continued)                         systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate; exceptions to this Specification are stated in the individual Specifications, or
: c.       When an allowance is stated in the individual value, parameter, or other Specification.
This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.
LCO 3.0.5           Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY Note that                    of other equipment. This is an exception to LCO 3.0.2 for the system returned to troubleshooting is            service under administrative control to perform the testing required to not allowed.                  demonstrate OPERABILITY.
Can Use this to retest:
: 1. inoperable equipment LCO 3.0.6           When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support OR system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, additional evaluations and limitations may be
: 2. other equipment.
required in accordance with Specification 5.7.2.18, "Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.
(continued)
Watts Bar-Unit 1                                3.0-2                                      Amendment 55


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 3 of 16  II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]: A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)] Some examples of SRO exam items for this topic include:
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]:
A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)]
Some examples of SRO exam items for this topic include:
* Reporting requirements when the maximum licensed thermal power output is exceeded.
* Reporting requirements when the maximum licensed thermal power output is exceeded.
* Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems, fire doors, etc.
* Administration of fire protection program requirements such as compensatory actions associated with inoperable sprinkler systems, fire doors, etc.
* The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
* The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
* National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
* National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
* Processes for TS and FSAR changes. Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic. B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)] Some examples of SRO exam items for this topic include:
* Processes for TS and FSAR changes.
Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic.
B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]
Some examples of SRO exam items for this topic include:
* Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
* Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
* Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
* Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
* Knowledge of TS bases that are required to analyze TS required actions and terminology.
* Knowledge of TS bases that are required to analyze TS required actions and terminology.
* Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).
* Same items listed above for the Technical Requirements Manual (TRM) and Offsite Dose Calculation Manual (ODCM).
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of  1 hour action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.  
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of  1 hour action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.
SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. ROs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Page 3 of 16


SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. RO's are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 4 of 16 ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)         Above this line RO knowledge Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 5 of 16 Figure 1:  Screening for SRO-only linked to 10 CFR 55.43(b)(2) (Tech Specs)       
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)
RO knowledge Above this line Page 4 of 16


Can question be answered solely by knowing  1 hour TS/TRM Action? RO question YesNoCan question be answered solely by knowing the LCO/TRM information listed "above-the-line?" YesRO question NoCan question be answered solely by knowing the TS Safety Limits? YesRO question NoDoes the question involve one or more of the following for TS, TRM, or ODCM?
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)
(Tech Specs)
Can question be answered solely by knowing  1       Yes hour TS/TRM Action?                                           RO question No Can question be answered solely by knowing the       Yes LCO/TRM information listed above-the-line?                   RO question No Can question be answered solely by knowing the       Yes TS Safety Limits?                                             RO question No Does the question involve one or more of the following for TS, TRM, or ODCM?
* Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
* Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)
* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)                             Yes      SRO-only
* Knowledge of TS bases that is required to analyze TS required actions and terminology SRO-only question YesNoQuestion might not be linked to 10 CFR 55.43(b)(2) for SRO-only
* Knowledge of TS bases that is required to analyze TS                   question required actions and terminology No Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only Page 5 of 16
: 96. In accordance with NPG-SPP-01.2, Administration of Site Technical Procedures, which ONE of the following describes the requirements for making a MINOR/EDITORIAL change to a technical procedure? An Independent Qualified Review (IQR) ____(1)____ REQUIRED. A 50.59 Screening Review ____(2)____REQUIRED. A. (1) is (2) is B. (1) is (2) is NOT C. (1) is NOT (2) is D. (1) is NOT (2) is NOT CORRECT ANSWER:B DISTRACTOR ANALYSIS:   A. Incorrect: NPG-SPP-01.2, Administration of Site Technical Procedures, 3.2.11.A describes that an AOR is not required but an IQR is and a 50.59 screening is not required. 50.59 screening requirements are directed by step 3.2.9.J   B. Correct: NPG-SPP-01.2, Administration of Site Technical Procedures, 3.2.11.A describes how minor editorial changes do require an IQR to be performed but a 50.59 screening is not required. C. Incorrect: NPG-SPP-01.2, Administration of Site Technical Procedures, 3.2.11.A describes that an AOR is not required but an IQR is 50.59 screening requirements are directed by step 3.2.9.J   D. Incorrect: Administration of Site Technical Procedures, 3.2.11.A describes how minor editorial changes do require an IQR to be performed a 50.59 screening is not required. 50.59 screening requirements are directed by step 3.2.9.J Question Number: 96  Tier:  3 Group:
 
K/A: G2.2 Equipment Control 2.2.6  Knowledge of the process for making changes to procedures  Importance Rating:  3.0  3.6 10 CFR Part 55: 
96.
In accordance with NPG-SPP-01.2, Administration of Site Technical Procedures, which ONE of the following describes the requirements for making a MINOR/EDITORIAL change to a technical procedure?
An Independent Qualified Review (IQR) ____(1)____ REQUIRED.
A 50.59 Screening Review ____(2)____REQUIRED.
A.   (1)   is (2)   is B.   (1)   is (2)   is NOT C.   (1)   is NOT (2)   is D.   (1)   is NOT (2)   is NOT
 
CORRECT ANSWER:                                                           B DISTRACTOR ANALYSIS:
A. Incorrect: NPG-SPP-01.2, Administration of Site Technical Procedures, 3.2.11.A describes that an AOR is not required but an IQR is and a 50.59 screening is not required. 50.59 screening requirements are directed by step 3.2.9.J B. Correct: NPG-SPP-01.2, Administration of Site Technical Procedures, 3.2.11.A describes how minor editorial changes do require an IQR to be performed but a 50.59 screening is not required.
C. Incorrect: NPG-SPP-01.2, Administration of Site Technical Procedures, 3.2.11.A describes that an AOR is not required but an IQR is 50.59 screening requirements are directed by step 3.2.9.J D. Incorrect: Administration of Site Technical Procedures, 3.2.11.A describes how minor editorial changes do require an IQR to be performed a 50.59 screening is not required. 50.59 screening requirements are directed by step 3.2.9.J


10CFR55.43.b: 10 CFR 55.43(b)(3) K/A Match: K/A is matched because the applicant is required to demonstrate the knowledge of which reviews are required for a procedure change by the processes of the facility.
Question Number:        96 Tier:    3  Group:
K/A:    G2.2 Equipment Control 2.2.6 Knowledge of the process for making changes to procedures Importance Rating:      3.0 3.6 10 CFR Part 55:
10CFR55.43.b:           10 CFR 55.43(b)(3)
K/A Match:   K/A is matched because the applicant is required to demonstrate the knowledge of which reviews are required for a procedure change by the processes of the facility.
Technical  
Technical  


==Reference:==
==Reference:==
NPG-SPP-01.2, Administration of Site Technical Procedures Proposed references to be provided: None  Learning Objective: 3-OT-AdminWB, NPG-SPP-01.02 Workbook 1-10 Cognitive Level:     Higher   Lower X   Question Source:     New   Modified Bank   Bank X   Question History: Bank question G 2.2.6 97, last used on the 06/2011 NRC Exam. Comments:  
NPG-SPP-01.2, Administration of Site Technical Procedures Proposed references to       None be provided:
Learning Objective:         3-OT-AdminWB, NPG-SPP-01.02 Workbook 1-10 Cognitive Level:
Higher Lower                 X Question Source:
New Modified Bank Bank                 X Question History:           Bank question G 2.2.6 97, last used on the 06/2011 NRC Exam.
Comments:


NPG Standard Programs and Processes Administration of Site Technical Procedures NPG-SPP-01.2 Rev. 0011 Page 16 of 54 3.2.9 Procedure Review Requirements (continued)     B. Procedures changing a QC holdpoint require an AOR by Quality Assurance. C. A review for incorporation of NQAP requirements must be performed by Quality Assurance personnel or others knowledgeable of the QA requirements. This review will typically be performed as part of the Independent Qualified Review (IQR). 1. Quality Related procedures require technical adequacy review by an Independent Qualified Review (IQR). Form TVA 40667 (NPG-SPP-01.2-3, Procedure Verification Review Checklist) shall be used by the IQR for the review. IQR reviewers shall not be the person who prepared the procedure. 2. The responsible department manager shall select individuals to participate in the IQR program and these individuals shall complete site IQR training. During the procedure review process, the IQR reviewer shall identify any additional cross disciplinary review required to the procedure writer. See Section 3.2.24 for the qualification requirements for IQR. 3. When extensive changes that have been made (for example, 50% or more of the procedures) and technical change of content, a full-scope IQR is required. D. PORC Review Required (per Technical Specification/NQAP). PORC review is required for the following:   1. New procedures or changes to existing procedures recommended by Regulatory Guide 1.33, Revision 2, Appendix A, February 1978; that require an evaluation in accordance with 10 CFR 50.59. 2. The emergency operating procedures which implement NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33. 3. Physical Security Plan.  
NPG Standard             Administration of Site Technical           NPG-SPP-01.2 Programs and                        Procedures                      Rev. 0011 Processes                                                          Page 16 of 54 3.2.9   Procedure Review Requirements (continued)
: 4. Radiological Emergency Plan.  
B. Procedures changing a QC holdpoint require an AOR by Quality Assurance.
: 5. Offsite Dose Calculation Manual. 6. Process Control Program (radwaste packaging and shipping). 7. Additional PORC reviews specifically required by site specific technical specifications or the plant's licensing basis. 8. Proposed changes to TS; Technical Requirements Manual; their bases; amendments to the Operating License. 9. Selected 10 CFR 50.59 evaluations. 10. Selected 10 CFR 72.48 evaluations.
C. A review for incorporation of NQAP requirements must be performed by Quality Assurance personnel or others knowledgeable of the QA requirements. This review will typically be performed as part of the Independent Qualified Review (IQR).
NPG Standard Programs and Processes Administration of Site Technical Procedures NPG-SPP-01.2 Rev. 0011 Page 18 of 54 3.2.9 Procedure Review Requirements (continued)     I. New technical procedures or changes to technical procedures that create or revise TEMPORARY MODIFICATIONS to structures, systems, and components (SSCs) for the purpose of utilizing the SSC for plant operation or crediting the SSC for performing a plant function shall be evaluated in accordance with NPG-SPP-09.5, Temporary Modifications, and shall be reviewed in accordance with NPG-SPP-09.4, 10 CFR 50.59 Evaluations of Changes, Tests, and Experiments and Step 3.2.9J and step 3.2.9L, as applicable. This does not include alterations for the purpose of testing the affected SSC. J. New technical procedures and changes to technical procedures shall be reviewed to determine if the procedure is within the scope of 10 CFR 50.59 using Attachment 1 of NPG-SPP-09.4. The results of this determination shall be noted on the PCF (Form TVA 40665 - NPG-SPP-01.2-1) or BSL/ECM audit trail. This review shall be performed by a 10 CFR 50.59 qualified individual in accordance with NPG-SPP-09.4, 10 CFR 50.59 Evaluations of Changes, Tests, and Experiments. Procedures shall be evaluated for 10 CFR 50.59 applicability if those procedures contain information described in the UFSAR such as how structures, systems, and components are operated and controlled, including assumed operator actions and response times. 1. If it is determined that 10 CFR 50.59 is applicable to the procedure or the change being made, then a 10 CFR 50.59 screening review shall be performed in accordance with NPG-SPP-09.4 using Forms TVA 40518 (NPG-SPP-09.4-1, Applicability Determination/Screening Review/50.59 Evaluation Coversheet) and TVA 40673 (NPG-SPP-09.4-2, Screening Review Form). Per NPG-SPP-09.4, Form TVA 40517 (NPG-SPP-09.4-7, Procedure Change Evaluation) may be used if appropriate. 2. If, as the result of the 10 CFR 50.59 screening review, a 10 CFR 50.59 evaluation needs to be generated, ensure the evaluation is performed by a qualified reviewer in accordance with NPG-SPP-09.4, 10 CFR 50.59 Evaluations of Changes, Tests, and Experiments. K. Changes to the Physical Security Contingency Plan, Radiological Emergency Plan and Implementing Procedures, and the NQAP do not require 10 CFR 50.59 screening reviews. Changes to these documents are made in accordance with 10 CFR 50.54, Conditions of Licenses. NOTE Minor/editorial changes do not require 10 CFR 72.48 reviews. The 10 CFR 72.48 documents will be archived in EDM as stand alone documents. L. New technical procedures or changes to technical procedures associated with Independent Spent Fuel Storage Installation (ISFSI) or shared (interfacing) systems which may impact ISFSI shall be reviewed to determine if the procedure or the change is within the scope of 10 CFR 72.48 in accordance with NPG-SPP-09.9, 10 CFR 72.48 Evaluations of Changes, Tests, and Experiments for Independent Spent Fuel Storage Installation.
: 1. Quality Related procedures require technical adequacy review by an Independent Qualified Review (IQR). Form TVA 40667 (NPG-SPP-01.2-3, Procedure Verification Review Checklist) shall be used by the IQR for the review. IQR reviewers shall not be the person who prepared the procedure.
NPG Standard Programs and Processes Administration of Site Technical Procedures NPG-SPP-01.2 Rev. 0011 Page 21 of 54   3.2.11 Minor/Editorial Changes A. Minor changes, such as inconsequential editorial corrections that do not affect the outcome, results, functions, processes, responsibilities, and requirements of the performance of procedure or instructions, require review by an IQR for quality-related procedures and approval by the appropriate approval authority. Minor changes do not require an AOR, 10 CFR 50.59 review, 10 CFR 72.48 review, or PORC review. Minor changes shall not change the intent of the procedure or alter the technical content or sequence of procedural steps. B. Procedure changes that meet any of the following criteria are considered minor changes: 1. Correction of punctuation, style changes 2. Redundant or insignificant word or title changes
: 2. The responsible department manager shall select individuals to participate in the IQR program and these individuals shall complete site IQR training. During the procedure review process, the IQR reviewer shall identify any additional cross disciplinary review required to the procedure writer. See Section 3.2.24 for the qualification requirements for IQR.
: 3. Correction of typographical errors including capitalization
: 3. When extensive changes that have been made (for example, 50% or more of the procedures) and technical change of content, a full-scope IQR is required.
: 4. Annotation of critical steps  
D. PORC Review Required (per Technical Specification/NQAP). PORC review is required for the following:
: 5. Correction of reference errors 6. Omitted symbols that do not alter results 7. Incorrect units of measure due to editorial error
: 1. New procedures or changes to existing procedures recommended by Regulatory Guide 1.33, Revision 2, Appendix A, February 1978; that require an evaluation in accordance with 10 CFR 50.59.
: 8. Misplaced decimals that are neither setpoint values nor tolerances
: 2. The emergency operating procedures which implement NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33.
: 9. Page number discrepancies
: 3. Physical Security Plan.
: 10. Missing sign-offs, signatures, or date lines
: 4. Radiological Emergency Plan.
: 11. Corrections to attachment identifiers 12. Corrections to titles of plant organizations, position titles, department/section/unit names when there is no change in authority, responsibility, or reporting relationships 13. Corrections to addresses, telephone numbers, or computer application names  
: 5. Offsite Dose Calculation Manual.
: 14. Corrections to or additions of equipment nomenclature or locations in procedures to be consistent with approved drawings, documents, labels, or procedure content 15. Addition of or changes to equipment unique identifier information (unid) in procedures consistent with design output documents and which do not alter what component is operated 16. Corrections to or clarification of a note or precaution which does not alter the method of accomplishing a task Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 6 of 16  C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)] Some examples of SRO exam items for this topic include:
: 6. Process Control Program (radwaste packaging and shipping).
: 7. Additional PORC reviews specifically required by site specific technical specifications or the plants licensing basis.
: 8. Proposed changes to TS; Technical Requirements Manual; their bases; amendments to the Operating License.
: 9. Selected 10 CFR 50.59 evaluations.
: 10. Selected 10 CFR 72.48 evaluations.
 
NPG Standard               Administration of Site Technical         NPG-SPP-01.2 Programs and                          Procedures                    Rev. 0011 Processes                                                        Page 18 of 54 3.2.9   Procedure Review Requirements (continued)
I. New technical procedures or changes to technical procedures that create or revise TEMPORARY MODIFICATIONS to structures, systems, and components (SSCs) for the purpose of utilizing the SSC for plant operation or crediting the SSC for performing a plant function shall be evaluated in accordance with NPG-SPP-09.5, Temporary Modifications, and shall be reviewed in accordance with NPG-SPP-09.4, 10 CFR 50.59 Evaluations of Changes, Tests, and Experiments and Step 3.2.9J and step 3.2.9L, as applicable. This does not include alterations for the purpose of testing the affected SSC.
J. New technical procedures and changes to technical procedures shall be reviewed to determine if the procedure is within the scope of 10 CFR 50.59 using Attachment 1 of NPG-SPP-09.4. The results of this determination shall be noted on the PCF (Form TVA 40665 - NPG-SPP-01.2-1) or BSL/ECM audit trail. This review shall be performed by a 10 CFR 50.59 qualified individual in accordance with NPG-SPP-09.4, 10 CFR 50.59 Evaluations of Changes, Tests, and Experiments. Procedures shall be evaluated for 10 CFR 50.59 applicability if those procedures contain information described in the UFSAR such as how structures, systems, and components are operated and controlled, including assumed operator actions and response times.
: 1. If it is determined that 10 CFR 50.59 is applicable to the procedure or the change being made, then a 10 CFR 50.59 screening review shall be performed in accordance with NPG-SPP-09.4 using Forms TVA 40518 (NPG-SPP-09.4-1, Applicability Determination/Screening Review/50.59 Evaluation Coversheet) and TVA 40673 (NPG-SPP-09.4-2, Screening Review Form). Per NPG-SPP-09.4, Form TVA 40517 (NPG-SPP-09.4-7, Procedure Change Evaluation) may be used if appropriate.
: 2. If, as the result of the 10 CFR 50.59 screening review, a 10 CFR 50.59 evaluation needs to be generated, ensure the evaluation is performed by a qualified reviewer in accordance with NPG-SPP-09.4, 10 CFR 50.59 Evaluations of Changes, Tests, and Experiments.
K. Changes to the Physical Security Contingency Plan, Radiological Emergency Plan and Implementing Procedures, and the NQAP do not require 10 CFR 50.59 screening reviews. Changes to these documents are made in accordance with 10 CFR 50.54, Conditions of Licenses.
NOTE Minor/editorial changes do not require 10 CFR 72.48 reviews. The 10 CFR 72.48 documents will be archived in EDM as stand alone documents.
L. New technical procedures or changes to technical procedures associated with Independent Spent Fuel Storage Installation (ISFSI) or shared (interfacing) systems which may impact ISFSI shall be reviewed to determine if the procedure or the change is within the scope of 10 CFR 72.48 in accordance with NPG-SPP-09.9, 10 CFR 72.48 Evaluations of Changes, Tests, and Experiments for Independent Spent Fuel Storage Installation.
 
NPG Standard           Administration of Site Technical             NPG-SPP-01.2 Programs and                      Procedures                      Rev. 0011 Processes                                                        Page 21 of 54 3.2.11   Minor/Editorial Changes A. Minor changes, such as inconsequential editorial corrections that do not affect the outcome, results, functions, processes, responsibilities, and requirements of the performance of procedure or instructions, require review by an IQR for quality-related procedures and approval by the appropriate approval authority. Minor changes do not require an AOR, 10 CFR 50.59 review, 10 CFR 72.48 review, or PORC review. Minor changes shall not change the intent of the procedure or alter the technical content or sequence of procedural steps.
B. Procedure changes that meet any of the following criteria are considered minor changes:
: 1. Correction of punctuation, style changes
: 2. Redundant or insignificant word or title changes
: 3. Correction of typographical errors including capitalization
: 4. Annotation of critical steps
: 5. Correction of reference errors
: 6. Omitted symbols that do not alter results
: 7. Incorrect units of measure due to editorial error
: 8. Misplaced decimals that are neither setpoint values nor tolerances
: 9. Page number discrepancies
: 10. Missing sign-offs, signatures, or date lines
: 11. Corrections to attachment identifiers
: 12. Corrections to titles of plant organizations, position titles, department/section/unit names when there is no change in authority, responsibility, or reporting relationships
: 13. Corrections to addresses, telephone numbers, or computer application names
: 14. Corrections to or additions of equipment nomenclature or locations in procedures to be consistent with approved drawings, documents, labels, or procedure content
: 15. Addition of or changes to equipment unique identifier information (unid) in procedures consistent with design output documents and which do not alter what component is operated
: 16. Corrections to or clarification of a note or precaution which does not alter the method of accomplishing a task
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]
Some examples of SRO exam items for this topic include:
* 10 CFR 50.59 screening and evaluation processes.
* 10 CFR 50.59 screening and evaluation processes.
* Administrative processes for temporary modifications.
* Administrative processes for temporary modifications.
* Administrative processes for disabling annunciators.
* Administrative processes for disabling annunciators.
* Administrative processes for the installation of temporary instrumentation.
* Administrative processes for the installation of temporary instrumentation.
* Processes for changing the plant or plant procedures. Section IV provides an example of a satisfactory SRO-only question related to this topic. D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.       [10 CFR 55.43(b)(4)] Some examples of SRO exam items for this topic include:
* Processes for changing the plant or plant procedures.
Section IV provides an example of a satisfactory SRO-only question related to this topic.
D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
[10 CFR 55.43(b)(4)]
Some examples of SRO exam items for this topic include:
* Process for gaseous/liquid release approvals, i.e., release permits.
* Process for gaseous/liquid release approvals, i.e., release permits.
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures. Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits. SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc. E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedure's overall mitigative strategy or purpose.  
Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.
SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.
E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.
The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:
Page 6 of 16
 
97.
Which ONE of the following describes the Containment Access authorization in accordance with TI-12.07A, Containment Access Modes 1 - 4?
General access into either the containment or the annulus; may be authorized provided that the incore flux detectors are in their normal storage location inside the crane wall
_____(1)_____ AND tagged with a _____(2)_____.
A.    (1)    ONLY (2)    Hold Order B.    (1)    ONLY (2)    Caution Order C.    (1)    OR approximately ten feet below the bottom of the core limit in any core thimble (2)    Hold Order D.    (1)    OR approximately ten feet below the bottom of the core limit in any core thimble (2)    Caution Order
 
CORRECT ANSWER:                                                              C DISTRACTOR ANALYSIS:
A. Incorrect: As seen in TI-12.07A, The incore flux detectors are in either of the two approved storage locations, as described in TI-41, (Normal STORAGE in Crane Wall or approximately ten feet below bottom of core limit in any core thimble). A Hold Order must be issued to the RADIATION PROTECTION Shift Supervisor and in place on the incore detector drive motors to prevent operation while personnel are inside Containment or Annulus.
Note that a hold order uses Danger Tags to indicate the isolation points of the hold order.
It is plausible to believe this distractor is correct because one may recollect the one approved storage location but not the other.
B. Incorrect: It is incorrect and yet plausible that the incores only have one approved storage location. It is also incorrect that a caution tag would be used to secure such. It is plausible to believe this as in the vast majority of the references to tagging the incores, TI-12.07A is mute as to the type of tag used. It simply states that the incores are to be TAGGED.
C. Correct: There are two approved storage locations for the incores. Also, as aforementioned, a Danger Tag is used.
D. Incorrect: While it is correct that there are two approved storage locations for the incores, it is not correct and yet plausible that a caution tag would be used to secure the incore detectors.


The applicant's knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedure's content is required to correctly answer the written test item, for example: 
Question Number:        97 Tier:    3  Group:
: 97. Which ONE of the following describes the Containment Access authorization in accordance with TI-12.07A, Containment Access Modes 1 - 4? General access into either the containment or the annulus; may be authorized provided that the incore flux detectors are in their normal storage location inside the crane wall _____(1)_____ AND tagged with a _____(2)_____. A. (1) ONLY (2) Hold Order B. (1) ONLY (2) Caution Order C. (1) OR approximately ten feet below the bottom of the core limit in any core thimble (2) Hold Order D. (1) OR approximately ten feet below the bottom of the core limit in any core thimble (2) Caution Order CORRECT ANSWER:C DISTRACTOR ANALYSIS:   A. Incorrect: As seen in TI-12.07A, "The incore flux detectors are in either of the two approved storage locations, as described in TI-41, (Normal STORAGE in Crane Wall or approximately ten feet below bottom of core limit in any core thimble). A Hold Order must be issued to the RADIATION PROTECTION Shift Supervisor and in place on the incore detector drive motors to prevent operation while personnel are inside Containment or Annulus."
K/A:    2.3 Radiation Control 2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.
Importance Rating:     3.2 3.7 10 CFR Part 55:      (CFR: 41.12 / 45.9 / 45.10) 10CFR55.43.b:        10 CFR 55.43(b)(4)
K/A Match:   K/A is matched because the applicant is required to demonstrate one of the SRO duties applicable to containment entry.
Technical


Note that a hold order uses Danger Tags to indicate the isolation points of the hold order.  
==Reference:==
TI-12.07A, Containment Access Modes 1-4 Proposed references to      None be provided:
Learning Objective:        3-OT-TI-1207, Containment Access
: 12. Discuss the precaution associated specifically to an entry into the annulus and lower containment.
Cognitive Level:
Higher Lower                X Question Source:
New Modified Bank Bank                X Question History:          Bank question G 2.3.12 97 used on the 09/2010 Sequoyah NRC exam.
Comments:


It is plausible to believe this distractor is correct because one may recollect the one approved storage location but not the other. B. Incorrect: It is incorrect and yet plausible that the incores only have one approved storage location. It is also incorrect that a caution tag would be used to secure such. It is plausible to believe this as in the vast majority of the references to tagging the incores, TI-12.07A is mute as to the type of tag used. It simply states that the incores are to be TAGGED. C. Correct: There are two approved storage locations for the incores. Also, as aforementioned, a Danger Tag is used. D. Incorrect: While it is correct that there are two approved storage locations for the incores, it is not correct and yet plausible that a caution tag would be used to secure the incore detectors.
WBN                    Containment Access                  TI-12.07A Unit 1                      Modes 1 - 4                    Rev. 0007 Page 17 of 50 3.2.3    Operations A. The Shift Manager (SM) or designee (SRO) reviews and authorizes all General Access entries into Containment or Annulus by completion of Appendix A Section 1.0, Authorization for General Access, after ensuring the following:
Question Number: 97  Tier:  3 Group:
: 1. The incore flux detectors are in either of the two approved storage locations, as described in TI-41, (Normal STORAGE in Crane Wall or approximately ten feet below bottom of core limit in any core thimble). A Hold Order must be issued to the RADIATION PROTECTION Shift A hold order is a          Supervisor and in place on the incore detector drive motors to prevent danger tag.                 operation while personnel are inside Containment or Annulus.[C5,6]
K/A: 2.3 Radiation Control 2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.
: 2. Access Control Custodian established OR airlock door alarms are enabled.
Importance Rating: 3.2 3.7 10 CFR Part 55: (CFR: 41.12 / 45.9 / 45.10) 10CFR55.43.b: 10 CFR 55.43(b)(4)
: a. The requirement for establishing an Access Control Custodian may be waived by the SM in the event that urgent entry is required. In these cases the airlock door alarm will remain enabled.
: 3. The work activity or evolution is approved to be performed, including authorization of entries inside the Polar Crane Wall when below Mode 2.
B. IF personnel require access inside the Polar Crane Wall while in Mode 1 or 2 OR require entry when the incore detectors are NOT TAGGED or are NOT properly stored, the SM (concurrent with the RP Manager) must evaluate the necessity of the entry, issue special instructions (if any), and authorize the entry by completion of Appendix A, Section 2.0 Authorization for Special Access.
Such entries require issuance of a special ALARA Plan and approvals in accordance RCI-128, ALARA Program. [C5,6]
C. WHEN the radiological hazard associated with a Special Access entry is no longer present, THEN the Shift Manager (concurrent with the RP Manager) may relax the entry requirements allowing the return to General Access by the completion of Appendix A, Section 3.0, Exit From Special Access Requirements.
D. The Shift Manager ensures that in the event of an evacuation from Upper Containment through Lower Containment that the incore detectors are placed in a safe condition (tagged or stored) prior to authorizing personnel to open the Personnel Hatch #2 (Subhatch, 757) for exit from Containment.
E. The Access Control Custodian will be briefed on responsibilities and expectations for the implementation of this TI. The briefing as a minimum is to cover items contained in Step 3.2.1. C and D.


K/A Match: K/A is matched because the applicant is required to demonstrate one of the SRO duties applicable to containment entry. Technical
WBN                      Containment Access                TI-12.07A Unit 1                        Modes 1 - 4                  Rev. 0007 Page 7 of 50 2.2  Developmental References (continued)
I. RCI-128, ALARA Program Implementation J. TI-134, Control of Portable Two Way Radios K. TI-229, Temporary Shielding Program L. WBN FSAR Questions 22.26, 212.116, and 212.129 3.0  PRECAUTIONS AND LIMITATIONS 3.1  General Precautions and Limitations A. All access portals (Airlocks el 757/716, equipment hatches el 757, and Annulus el 713) to the Containment building SHALL be controlled to prevent unauthorized entry while in Modes 1 through 4. Radiation Protection (RP) shall maintain positive access control of Containment and Annulus in accordance with RP procedures.
B. All entries into Containment or Annulus while in Modes 1 through 4 are authorized by completion of applicable sections of Appendix A, Containment/Annulus Entry Authorization.
C. One Appendix A is required for each area entered (Upper Containment, Lower Containment, Annulus). Appendix As are typically issued for a 24 hour period only. However the SM may approve an extension beyond 24 hours.
Authorization requirements for entries are described as follows:
: 1. General Access is authorized by the completion of Appendix A, Section 1.0, which requires the incore detectors to be TAGGED and in either of the two approved storage locations, as described in TI-41, (Normal STORAGE in Crane Wall or approximately ten feet below bottom of core limit in any core thimble). This ensures that a detector is in a location which would not expose entry personnel. In addition, this section also permits entry inside the Polar Crane Wall when below Mode 2.
: 2. Special Access is authorized by completion of Appendix A, Section 2.0, for any entry into Containment or Annulus when the incore detectors are NOT TAGGED or NOT in their approved storage location. Special Access authorization is also required for entries inside the Polar Crane Wall or Regen Hx Room during Modes 1 or 2.
: 3. All entries requiring Special Access Authorization shall require approvals in accordance with RCI-128, ALARA Program Implementation to ensure that appropriate controls are established to prevent personnel overexposure (i.e., Special RWP and APR).


==Reference:==
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
TI-12.07A, Containment Access Modes 1-4 Proposed references to be provided: None  Learning Objective: 3-OT-TI-1207, Containment Access 12. Discuss the precaution associated specifically to an entry into the annulus and lower containment. Cognitive Level:      Higher  Lower X  Question Source:      New  Modified Bank  Bank X  Question History: Bank question G 2.3.12 97 used on the 09/2010 Sequoyah NRC exam. Comments:
C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]
WBN Unit 1 Containment Access Modes 1 - 4 TI-12.07A Rev. 0007 Page 17 of 50  3.2.3 Operations  A. The Shift Manager (SM) or designee (SRO) reviews and authorizes all General Access entries into Containment or Annulus by completion of Appendix A Section 1.0, Authorization for General Access, after ensuring the following: 1. The incore flux detectors are in either of the two approved storage locations, as described in TI-41, (Normal STORAGE in Crane Wall or approximately ten feet below bottom of core limit in any core thimble). A Hold Order must be issued to the RADIATION PROTECTION Shift Supervisor and in place on the incore detector drive motors to prevent operation while personnel are inside Containment or Annulus.[C5,6]  2. Access Control Custodian established OR airlock door alarms are enabled. a. The requirement for establishing an Access Control Custodian may be  waived by the SM in the event that urgent entry is required. In these cases the airlock door alarm will remain enabled. 3. The work activity or evolution is approved to be performed, including authorization of entries inside the Polar Crane Wall when below Mode 2. B. IF personnel require access inside the Polar Crane Wall while in Mode 1 or 2 OR require entry when the incore detectors are NOT TAGGED or are NOT  properly stored, the SM (concurrent with the RP Manager) must evaluate the necessity of the entry, issue special instructions (if any), and authorize the entry by completion of Appendix A, Section 2.0 Authorization for Special Access. Such entries require issuance of a special ALARA Plan and approvals in accordance RCI-128, ALARA Program. [C5,6]    C. WHEN the radiological hazard associated with a Special Access entry is no longer present, THEN the Shift Manager (concurrent with the RP Manager) may relax the entry requirements allowing the return to General Access by the completion of Appendix A, Section 3.0, Exit From Special Access Requirements. D. The Shift Manager ensures that in the event of an evacuation from Upper Containment through Lower Containment that the incore detectors are placed in a safe condition (tagged or stored) prior to authorizing personnel to open the Personnel Hatch #2 (Subhatch, 757') for exit from Containment. E. The Access Control Custodian will be briefed on responsibilities  and expectations for the implementation of this TI. The briefing as a minimum is to cover items contained in Step 3.2.1. C and D.
Some examples of SRO exam items for this topic include:
WBN Unit 1 Containment Access Modes 1 - 4 TI-12.07A Rev. 0007 Page 7 of 50  2.2 Developmental References (continued)    I. RCI-128, ALARA Program Implementation J. TI-134, Control of Portable Two Way Radios K. TI-229, Temporary Shielding Program L. WBN FSAR Questions 22.26, 212.116, and 212.129  3.0 PRECAUTIONS AND LIMITATIONS 3.1 General Precautions and Limitations  A. All access portals (Airlocks el 757/716, equipment hatches el 757, and Annulus el 713) to the Containment building SHALL be controlled to prevent unauthorized entry while in Modes 1 through 4. Radiation Protection (RP) shall maintain positive access control of Containment and Annulus in accordance with RP procedures. B. All entries into Containment or Annulus while in Modes 1 through 4 are authorized by completion of applicable sections of Appendix A, Containment/Annulus Entry Authorization. C. One Appendix A is required for each area entered (Upper Containment, Lower Containment, Annulus). Appendix A's are typically issued for a 24 hour period only. However the SM may approve an extension beyond 24 hours. Authorization requirements for entries are described as follows: 1. General Access is authorized by the completion of Appendix A, Section 1.0, which requires the incore detectors to be TAGGED and in either of the two approved storage locations, as described in TI-41, (Normal STORAGE in Crane Wall or approximately ten feet below bottom of core limit in any core thimble). This ensures that a detector is in a location which would not expose entry personnel. In addition, this section also permits entry inside the Polar Crane Wall when below Mode 2. 2. Special Access is authorized by completion of Appendix A, Section 2.0, for any entry into Containment or Annulus when the incore detectors are NOT TAGGED or NOT in their approved storage location. Special Access authorization is also required for entries inside the Polar Crane Wall or Regen Hx Room during Modes 1 or 2. 3. All entries requiring Special Access Authorization shall require approvals in accordance with RCI-128, ALARA Program Implementation to ensure that appropriate controls are established to prevent personnel overexposure (i.e., Special RWP and APR).
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 6 of 16  C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)] Some examples of SRO exam items for this topic include:
* 10 CFR 50.59 screening and evaluation processes.
* 10 CFR 50.59 screening and evaluation processes.
* Administrative processes for temporary modifications.
* Administrative processes for temporary modifications.
* Administrative processes for disabling annunciators.
* Administrative processes for disabling annunciators.
* Administrative processes for the installation of temporary instrumentation.
* Administrative processes for the installation of temporary instrumentation.
* Processes for changing the plant or plant procedures. Section IV provides an example of a satisfactory SRO-only question related to this topic. D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.       [10 CFR 55.43(b)(4)] Some examples of SRO exam items for this topic include:
* Processes for changing the plant or plant procedures.
Section IV provides an example of a satisfactory SRO-only question related to this topic.
D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
[10 CFR 55.43(b)(4)]
Some examples of SRO exam items for this topic include:
The requirement
* Process for gaseous/liquid release approvals, i.e., release permits.
* Process for gaseous/liquid release approvals, i.e., release permits.
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures. Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits. SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc. E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
tested in this
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedure's overall mitigative strategy or purpose.  
* Analysis and interpretation of radiation and activity readings as they question is one in        pertain to selection of administrative, normal, abnormal, and emergency which ONLY a SRO          procedures.
can sign for. Also,      Analysis and interpretation of coolant activity, including comparison to the SRO is the one        emergency plan criteria and/or regulatory limits.
who will sign for (in eSOMS) the holder    SRO-only knowledge should not be claimed for questions that can be of the clearance. answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.
E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.
The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:
Page 6 of 16


The applicant's knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedure's content is required to correctly answer the written test item, for example: 
98.
: 98. Given the following conditions: - Unit 1 is at 100% power. - A Containment entry INSIDE of the Polar Crane wall is REQUIRED. Which ONE of the following describes the Containment entry in accordance with TI-12.07A, Containment Access Modes 1-4? ____(1)____ can approve the containment entry listed above. The personnel hazard stipulated by TI-12.07A can be mitigated by the crew reducing ____(2)____. A. (1) ONLY the SM (2) reactor power in accordance with 1-GO-4 B. (1) ONLY the SM (2) containment temperature in accordance with SOI-30.02 C. (1) EITHER the SM or the US (2) reactor power in accordance with 1-GO-4 D. (1) EITHER the SM or the US (2) containment temperature in accordance with SOI-30.02   NOTE: 1-GO-4, Normal Power Operation  section 5.3 of 1-GO-4,  5.3 Unit Shutdown from 100% to 30% Reactor Power SOI-30.02, Containment Purge System CORRECT ANSWER:A DISTRACTOR ANALYSIS:  A. Correct: As seen in section 3.2.3 of TI-12.07A, it is true that "the Shift Manager (SM) or designee (SRO) reviews and authorizes all General Access entries into Containment-IF personnel require access inside the Polar Crane Wall while in Mode 1 or 2-the SM (concurrent with the RP Manager)-authorize the entry by completion of Appendix A, Section 2.0 Authorization for Special Access."  Note that Appendix A (page 2 of 3) does not allow a designee to sign for the shift manager (thus authorizing the special access to containment).
Given the following conditions:
TI-12.07A lists that "All entries requiring Special Access Authorization shall require approvals-to ensure that appropriate controls are established to prevent personnel overexposure."  Therefore, the concern at hand with respect to TI-12.07A is radiological dose. Radiological dose can be reduced by reducing reactor power. Therefore it is correct that the concern explicated in TI-12.07A may be mitigated by reducing power using 1-GO-4. B. Incorrect: As described, the SM provides the approval for the containment entry. It is not correct that the crew would purge the containment to mitigate the concern of TI-12.07A as this TI indicates that the concern at hand is personnel overexposure. 
    - Unit 1 is at 100% power.
    - A Containment entry INSIDE of the Polar Crane wall is REQUIRED.
Which ONE of the following describes the Containment entry in accordance with TI-12.07A, Containment Access Modes 1-4?
____(1)____ can approve the containment entry listed above.
The personnel hazard stipulated by TI-12.07A can be mitigated by the crew reducing
____(2)____.
NOTE:      1-GO-4, Normal Power Operation section 5.3 of 1-GO-4, 5.3 Unit Shutdown from 100% to 30% Reactor Power SOI-30.02, Containment Purge System A.   (1)     ONLY the SM (2)     reactor power in accordance with 1-GO-4 B.   (1)     ONLY the SM (2)     containment temperature in accordance with SOI-30.02 C.   (1)     EITHER the SM or the US (2)     reactor power in accordance with 1-GO-4 D.   (1)     EITHER the SM or the US (2)     containment temperature in accordance with SOI-30.02


By practical experience, lower containment is hot. By the Unit T/S, lower containment is over 100&deg;F (when in Mode 1). Because of this, one must be concerned with heat stress and stay time calculations. Inevitably what would be done to correct this would be to run containment purge. However, this would be done to satisfy the concerns of the safety manual procedure, TVA-TSP-18.906, "Heat Stress."  TI-12.07A does not contain any reference to the temperature of containment.  
CORRECT ANSWER:                                                                A DISTRACTOR ANALYSIS:
A. Correct: As seen in section 3.2.3 of TI-12.07A, it is true that the Shift Manager (SM) or designee (SRO) reviews and authorizes all General Access entries into ContainmentIF personnel require access inside the Polar Crane Wall while in Mode 1 or 2the SM (concurrent with the RP Manager)authorize the entry by completion of Appendix A, Section 2.0 Authorization for Special Access. Note that Appendix A (page 2 of 3) does not allow a designee to sign for the shift manager (thus authorizing the special access to containment).
TI-12.07A lists that All entries requiring Special Access Authorization shall require approvalsto ensure that appropriate controls are established to prevent personnel overexposure. Therefore, the concern at hand with respect to TI-12.07A is radiological dose.
Radiological dose can be reduced by reducing reactor power.
Therefore it is correct that the concern explicated in TI-12.07A may be mitigated by reducing power using 1-GO-4.
B. Incorrect: As described, the SM provides the approval for the containment entry.
It is not correct that the crew would purge the containment to mitigate the concern of TI-12.07A as this TI indicates that the concern at hand is personnel overexposure.
By practical experience, lower containment is hot. By the Unit T/S, lower containment is over 100&deg;F (when in Mode 1). Because of this, one must be concerned with heat stress and stay time calculations.
Inevitably what would be done to correct this would be to run containment purge. However, this would be done to satisfy the concerns of the safety manual procedure, TVA-TSP-18.906, Heat Stress. TI-12.07A does not contain any reference to the temperature of containment.
It is plausible to believe that a purge would be conducted because such would be the case; again this would be done to satisfy the heat stress requirements found in the mentioned chapter of the safety manual and not those of TI-12.07A. Also amplifying this plausibility is the fact that TI-12.07B, Containment Access Modes 5&6 does regard heat stress as a concern and ensures that persons conducting closeout inspections of the containment (for entry into Mode 4) are briefed on heat stress.
C. Incorrect: While it is correct that a power reduction would mitigate the stated concern of TI-12.07A for the containment entry, it is not correct that either the SM or the US could authorize the entry. As discussed, only the SM can approve a special containment entry. It is plausible to believe this as the US can sign for a general containment access.
D. Incorrect: As previously discussed, it is incorrect and yet plausible that a


It is plausible to believe that a purge would be conducted because such would be the case; again this would be done to satisfy the heat stress requirements found in the mentioned chapter of the safety manual and not those of TI-12.07A. Also amplifying this plausibility is the fact that TI-12.07B, "Containment Access Modes 5&6" does regard heat stress as a concern and ensures that persons conducting closeout inspections of the containment (for entry into Mode 4) are briefed on heat stress. C. Incorrect: While it is correct that a power reduction would mitigate the stated concern of TI-12.07A for the containment entry, it is not correct that either the SM or the US could authorize the entry. As discussed, only the SM can approve a special containment entry. It is plausible to believe this as the US can sign for a general containment access. D. Incorrect: As previously discussed, it is incorrect and yet plausible that a containment purge would mitigate the concern of TI-12.07A. Also, it is incorrect and yet plausible that either the SM or the US could authorize the entry.
containment purge would mitigate the concern of TI-12.07A. Also, it is incorrect and yet plausible that either the SM or the US could authorize the entry.
Question Number: 98  Tier:  3 Group:
K/A: 2.3 Radiation Control 2.3.14 Knowledge of radiation or contamination hazards that may arise during normal,l abnormal, or emergency conditions or activities. Importance Rating: 3.4  3.8


10 CFR Part 55: (CFR: 41.12 / 43.4 / 45.10) 10CFR55.43.b: 10 CFR 55.43(b)(4) and 10 CFR 55.43(b)(5)
Question Number:        98 Tier:    3  Group:
K/A Match: K/A is matched because the applicant is required to exhibit knowledge of the hazards present during a special containment entry and then compare such to those presented in TI-12.07A to arrive at a mitigation strategy. Technical  
K/A:    2.3 Radiation Control 2.3.14 Knowledge of radiation or contamination hazards that may arise during normal,l abnormal, or emergency conditions or activities.
Importance Rating:      3.4 3.8 10 CFR Part 55:       (CFR: 41.12 / 43.4 / 45.10) 10CFR55.43.b:         10 CFR 55.43(b)(4) and 10 CFR 55.43(b)(5)
K/A Match:   K/A is matched because the applicant is required to exhibit knowledge of the hazards present during a special containment entry and then compare such to those presented in TI-12.07A to arrive at a mitigation strategy.
Technical  


==Reference:==
==Reference:==
TI-12.07A, Containment Access Modes 1-4 TI-12.07B, Containment Access Modes 5-6 Proposed references to be provided: None  Learning Objective: 3-OT-TI-1207, Containment Access 12. Discuss the precaution associated specifically to an entry into the annulus and lower containment Cognitive Level:     Higher X Lower     Question Source:     New X Modified Bank   Bank     Question History: New question for the 2015-301 NRC SRO Exam Comments:
TI-12.07A, Containment Access Modes 1-4 TI-12.07B, Containment Access Modes 5-6 Proposed references to     None be provided:
WBN Unit 1 Containment Access Modes 1 - 4 TI-12.07A Rev. 0007 Page 7 of 50 2.2 Developmental References (continued)     I. RCI-128, ALARA Program Implementation J. TI-134, Control of Portable Two Way Radios K. TI-229, Temporary Shielding Program L. WBN FSAR Questions 22.26, 212.116, and 212.129 3.0 PRECAUTIONS AND LIMITATIONS 3.1 General Precautions and Limitations A. All access portals (Airlocks el 757/716, equipment hatches el 757, and Annulus el 713) to the Containment building SHALL be controlled to prevent unauthorized entry while in Modes 1 through 4. Radiation Protection (RP) shall maintain positive access control of Containment and Annulus in accordance with RP procedures. B. All entries into Containment or Annulus while in Modes 1 through 4 are authorized by completion of applicable sections of Appendix A, Containment/Annulus Entry Authorization. C. One Appendix A is required for each area entered (Upper Containment, Lower Containment, Annulus). Appendix A's are typically issued for a 24 hour period only. However the SM may approve an extension beyond 24 hours. Authorization requirements for entries are described as follows: 1. General Access is authorized by the completion of Appendix A, Section 1.0, which requires the incore detectors to be TAGGED and in either of the two approved storage locations, as described in TI-41, (Normal STORAGE in Crane Wall or approximately ten feet below bottom of core limit in any core thimble). This ensures that a detector is in a location which would not expose entry personnel. In addition, this section also permits entry inside the Polar Crane Wall when below Mode 2. 2. Special Access is authorized by completion of Appendix A, Section 2.0, for any entry into Containment or Annulus when the incore detectors are NOT TAGGED or NOT in their approved storage location. Special Access authorization is also required for entries inside the Polar Crane Wall or Regen Hx Room during Modes 1 or 2. 3. All entries requiring Special Access Authorization shall require approvals in accordance with RCI-128, ALARA Program Implementation to ensure that appropriate controls are established to prevent personnel overexposure (i.e., Special RWP and APR).
Learning Objective:         3-OT-TI-1207, Containment Access
WBN Unit 1 Containment Access Modes 1 - 4 TI-12.07A Rev. 0007 Page 12 of 50   3.2.2 Personnel Entering/Exiting Containment A. Personnel needing to enter Containment or Annulus must request that RP initiate an Appendix A, to obtain authorization for entry. One Appendix A is required for each area entered (Upper Containment, Lower Containment, Annulus) and is typically issued for a 24 hour period. B. Personnel must notify RP Shift Supervision in advance of planned entries to coordinate support and establish access requirements. C. General entries into Containment or Annulus on days other than designated Containment days, requires the approval of the Shift Manager or designee and RP Superintendent or designee. The Work Week Manager should also be consulted to confirm the need for entry AND may be contingent on the following:
: 12. Discuss the precaution associated specifically to an entry into the annulus and lower containment Cognitive Level:
Higher               X Lower Question Source:
New                 X Modified Bank Bank Question History:           New question for the 2015-301 NRC SRO Exam Comments:
 
WBN                     Containment Access                 TI-12.07A Unit 1                        Modes 1 - 4                  Rev. 0007 Page 7 of 50 2.2     Developmental References (continued)
I. RCI-128, ALARA Program Implementation J. TI-134, Control of Portable Two Way Radios K. TI-229, Temporary Shielding Program L. WBN FSAR Questions 22.26, 212.116, and 212.129 3.0     PRECAUTIONS AND LIMITATIONS 3.1     General Precautions and Limitations A. All access portals (Airlocks el 757/716, equipment hatches el 757, and Annulus el 713) to the Containment building SHALL be controlled to prevent unauthorized entry while in Modes 1 through 4. Radiation Protection (RP) shall maintain positive access control of Containment and Annulus in accordance with RP procedures.
B. All entries into Containment or Annulus while in Modes 1 through 4 are authorized by completion of applicable sections of Appendix A, Containment/Annulus Entry Authorization.
C. One Appendix A is required for each area entered (Upper Containment, Lower Containment, Annulus). Appendix As are typically issued for a 24 hour period only. However the SM may approve an extension beyond 24 hours.
Authorization requirements for entries are described as follows:
: 1. General Access is authorized by the completion of Appendix A, Section 1.0, which requires the incore detectors to be TAGGED and in either of the two approved storage locations, as described in TI-41, (Normal STORAGE in Crane Wall or approximately ten feet below bottom of core limit in any core thimble). This ensures that a detector is in a location which would not expose entry personnel. In addition, this section also permits entry inside the Polar Crane Wall when below Mode 2.
: 2. Special Access is authorized by completion of Appendix A, Section 2.0, for any entry into Containment or Annulus when the incore detectors are NOT This represents the          TAGGED or NOT in their approved storage location. Special Access radiological hazard          authorization is also required for entries inside the Polar Crane Wall or of the special              Regen Hx Room during Modes 1 or 2.
access - personnel overexposure.          3. All entries requiring Special Access Authorization shall require approvals in accordance with RCI-128, ALARA Program Implementation to ensure that appropriate controls are established to prevent personnel overexposure (i.e., Special RWP and APR).
 
WBN                       Containment Access                 TI-12.07A Unit 1                          Modes 1 - 4                  Rev. 0007 Page 12 of 50 3.2.2   Personnel Entering/Exiting Containment A. Personnel needing to enter Containment or Annulus must request that RP initiate an Appendix A, to obtain authorization for entry. One Appendix A is required for each area entered (Upper Containment, Lower Containment, Annulus) and is typically issued for a 24 hour period.
B. Personnel must notify RP Shift Supervision in advance of planned entries to coordinate support and establish access requirements.
C. General entries into Containment or Annulus on days other than designated Containment days, requires the approval of the Shift Manager or designee and RP Superintendent or designee. The Work Week Manager should also be consulted to confirm the need for entry AND may be contingent on the following:
* Activity was scheduled prior to T-0 to be performed on requested day, OR
* Activity was scheduled prior to T-0 to be performed on requested day, OR
* Activity is emergent and High Priority, i.e., LCO, WO priority 1 or 2, Ops concern, Management concern, etc. AND
* Activity is emergent and High Priority, i.e., LCO, WO priority 1 or 2, Ops concern, Management concern, etc.
AND
* Activity cannot be rescheduled to work on another Containment entry day due to in-progress surveillance, NRC late date, FEG week, etc. AND
* Activity cannot be rescheduled to work on another Containment entry day due to in-progress surveillance, NRC late date, FEG week, etc. AND
* Resources are available to support entry and work. D. The Access Control Custodian is responsible for ensuring completion and submittal of paperwork to Operations for closure. E. All personnel must obtain an RWP briefing prior to entering Containment or Annulus, including a briefing on the alternate evacuation route through Personnel Hatch #2 (Subhatch, 757') between Upper and Lower Containment.
* Resources are available to support entry and work.
If the Upper Airlock is inoperable and an alternate evacuation route must be taken, RP and Operations must be contacted to ensure the incore detectors are tagged and properly stored prior to opening Personnel Hatch #2 (Subhatch, 757') between Upper and Lower Containment. F. Upon initial entry into Containment or Annulus the airlock/access telephones are to be checked for proper operation and documented on Appendix B, Section 4.0, Airlock Phone Checks. G. Entries into Containment or Annulus require that personnel be accounted for at all times. Accountability is maintained utilizing Appendix B, Personnel Accountability Logsheet(s). These logs are maintained in the following manner:
D. The Access Control Custodian is responsible for ensuring completion and submittal of paperwork to Operations for closure.
WBN Unit 1 Containment Access Modes 1 - 4 TI-12.07A Rev. 0007 Page 17 of 50   3.2.3 Operations A. The Shift Manager (SM) or designee (SRO) reviews and authorizes all General Access entries into Containment or Annulus by completion of Appendix A Section 1.0, Authorization for General Access, after ensuring the following: 1. The incore flux detectors are in either of the two approved storage locations, as described in TI-41, (Normal STORAGE in Crane Wall or approximately ten feet below bottom of core limit in any core thimble). A Hold Order must be issued to the RADIATION PROTECTION Shift Supervisor and in place on the incore detector drive motors to prevent operation while personnel are inside Containment or Annulus.[C5,6] 2. Access Control Custodian established OR airlock door alarms are enabled. a. The requirement for establishing an Access Control Custodian may be waived by the SM in the event that urgent entry is required. In these cases the airlock door alarm will remain enabled. 3. The work activity or evolution is approved to be performed, including authorization of entries inside the Polar Crane Wall when below Mode 2. B. IF personnel require access inside the Polar Crane Wall while in Mode 1 or 2 OR require entry when the incore detectors are NOT TAGGED or are NOT properly stored, the SM (concurrent with the RP Manager) must evaluate the necessity of the entry, issue special instructions (if any), and authorize the entry by completion of Appendix A, Section 2.0 Authorization for Special Access. Such entries require issuance of a special ALARA Plan and approvals in accordance RCI-128, ALARA Program. [C5,6]   C. WHEN the radiological hazard associated with a Special Access entry is no longer present, THEN the Shift Manager (concurrent with the RP Manager) may relax the entry requirements allowing the return to General Access by the completion of Appendix A, Section 3.0, Exit From Special Access Requirements. D. The Shift Manager ensures that in the event of an evacuation from Upper Containment through Lower Containment that the incore detectors are placed in a safe condition (tagged or stored) prior to authorizing personnel to open the Personnel Hatch #2 (Subhatch, 757') for exit from Containment. E. The Access Control Custodian will be briefed on responsibilities and expectations for the implementation of this TI. The briefing as a minimum is to cover items contained in Step 3.2.1. C and D.
E. All personnel must obtain an RWP briefing prior to entering Containment or Annulus, including a briefing on the alternate evacuation route through Personnel Hatch #2 (Subhatch, 757) between Upper and Lower Containment.
WBN Unit 1 Containment Access Modes 1 - 4 TI-12.07A Rev. 0007 Page 31 of 50   Appendix A (Page 1 of 3) Containment/Annulus Entry Authorization   1.0 AUTHORIZATION FOR GENERAL ACCESS   Personnel Access To Be Used:(3)   757 Airlock (Upper)   716 Airlock (Lower)   713 Annulus Airlock Door Alarms Enabled?   YES Initials NO(4)   Initials Custodian Designated?   YES Initials NO(4)(6)   Initials Incore Detectors Tagged and Stored IAW TI-41 and NO entry inside the Polar Crane Wall while in Mode 2 or above. YES  Initials  Containment/Annulus Access Authorization: Mode(s): ____________   Entry Authorization:(1)    Unit SRO or SM Date Time Entry Authorization:(1)    RP Supt. or designee Date Time Remarks: ________________________________________________________________________________________  (1) RP Superintendent or designee and SM/SRO signature required for all Containment entries. (2) Entry inside the Polar Crane Wall or Regen Hx Room while in Modes 1 & 2 REQUIRES authorization by RP Manager AND SM in accordance with RCI-128 prior to entry. (3) IF Upper or Lower Containment entry required with incore probes NOT tagged or NOT Stored IAW TI-41, THEN Shift Manager AND RP Manager must authorize by completion of Section 2.0, Special Access Authorization.5, 6 Requirement for Upper Containment entry based on safety concern for Containment egress if upper airlock doors become inoperable. (4) Signature verifies availability of Access Control Custodian for Containment, if Airlock door alarms are not enabled. (5) Radiological hazard NO longer present - irradiated source NO longer poses potential for personnel exposure (Incore detectors TAGGED and properly stored OR removed) AND protective measures/controls are established. (6) The SM may waive establishing Access Custodian in the event of urgent entry.
If the Upper Airlock is inoperable and an alternate evacuation route must be taken, RP and Operations must be contacted to ensure the incore detectors are tagged and properly stored prior to opening Personnel Hatch #2 (Subhatch, 757) between Upper and Lower Containment.
WBN Unit 1 Containment Access Modes 1 - 4 TI-12.07A Rev. 0007 Page 32 of 50 Appendix A (Page 2 of 3) Containment/Annulus Entry Authorization   2.0 AUTHORIZATION FOR SPECIAL ACCESS   Personnel Access To Be Used:(3)   757 Airlock (Upper)   716 Airlock (Lower)   713 Annulus Airlock Door Alarms Enabled? YES Initials NO(4)   Initials Custodian Designated(3)   YES Initials     Authorized Work Group(s): ___________________________________________________________________   Description of Authorized Work: _______________________________________________________________ Work Limitations or Special Instructions (i.e., Access Control Requirements, specific ALARA Plan, etc.): ___________________________________________________________________________________________ ___________________________________________________________________________________________  Incore Detectors NOT Tagged or NOT Stored IAW TI-41(3)     Initials Polar Crane Wall (Modes  1 or 2) or Regen Hx Rm Entry Required(2)   Initials Containment/Annulus Access Authorization: Mode(s): ______________   Entry Authorization - Incore Detectors NOT TAGGED, NOT Stored IAW TI-41 OR Crane Wall entry required while in Mode 1 or 2.(2)(3)     Shift Manager Date Time Entry Authorization - Incore Detectors NOT TAGGED, NOT Stored IAW TI-41 OR Crane Wall entry required while in Mode 1 or 2.(2)(3)     RP Manager Date Time (1) RP Superintendent or designee and SM/SRO signature required for all Containment entries. (2) Entry inside the Polar Crane Wall or Regen Hx Room while in Modes 1 & 2 REQUIRES authorization by RP Manager AND SM in accordance with RCI-128 prior to entry. (3) IF Upper or Lower Containment entry required with incore probes NOT tagged or NOT Stored IAW TI-41, THEN Shift Manager AND RP Manager must authorize by completion of Section 2.0, Special Access Authorization.5, 6 Requirement for upper containment entry based on safety concern for containment egress if upper airlock doors become inoperable. (4) Signature verifies availability of Access Control Custodian for Containment, if Airlock door alarms are not enabled. (5) Radiological hazard NO longer present - irradiated source NO longer poses potential for personnel exposure (Incore detectors TAGGED and properly stored OR removed) AND protective measures/controls are established. (6) The SM may waive establishing Access Custodian in the event of urgent entry.
F. Upon initial entry into Containment or Annulus the airlock/access telephones are to be checked for proper operation and documented on Appendix B, Section 4.0, Airlock Phone Checks.
WBN Unit 0 Containment Access Modes 5 & 6 TI-12.07B Rev. 0006 Page 12 of 38         Date ________   4.1 Preliminary Actions (continued)     [8] ENSURE personnel who will be performing inspections are briefed on the following:
G. Entries into Containment or Annulus require that personnel be accounted for at all times. Accountability is maintained utilizing Appendix B, Personnel Accountability Logsheet(s). These logs are maintained in the following manner:
 
WBN                     Containment Access                   TI-12.07A Unit 1                      Modes 1 - 4                    Rev. 0007 Page 17 of 50 3.2.3   Operations A. The Shift Manager (SM) or designee (SRO) reviews and authorizes all General Access entries into Containment or Annulus by completion of Appendix A Section 1.0, Authorization for General Access, after ensuring the following:
: 1. The incore flux detectors are in either of the two approved storage locations, as described in TI-41, (Normal STORAGE in Crane Wall or approximately ten feet below bottom of core limit in any core thimble). A Hold Order must be issued to the RADIATION PROTECTION Shift Supervisor and in place on the incore detector drive motors to prevent operation while personnel are inside Containment or Annulus.[C5,6]
: 2. Access Control Custodian established OR airlock door alarms are enabled.
: a. The requirement for establishing an Access Control Custodian may be waived by the SM in the event that urgent entry is required. In these cases the airlock door alarm will remain enabled.
: 3. The work activity or evolution is approved to be performed, including authorization of entries inside the Polar Crane Wall when below Mode 2.
B. IF personnel require access inside the Polar Crane Wall while in Mode 1 or 2 OR require entry when the incore detectors are NOT TAGGED or are NOT properly stored, the SM (concurrent with the RP Manager) must evaluate the necessity of the entry, issue special instructions (if any), and authorize the entry by completion of Appendix A, Section 2.0 Authorization for Special Access.
Such entries require issuance of a special ALARA Plan and approvals in accordance RCI-128, ALARA Program. [C5,6]
C. WHEN the radiological hazard associated with a Special Access entry is no longer present, THEN the Shift Manager (concurrent with the RP Manager) may relax the entry requirements allowing the return to General Access by the completion of Appendix A, Section 3.0, Exit From Special Access Requirements.
D. The Shift Manager ensures that in the event of an evacuation from Upper Containment through Lower Containment that the incore detectors are placed in a safe condition (tagged or stored) prior to authorizing personnel to open the Personnel Hatch #2 (Subhatch, 757) for exit from Containment.
E. The Access Control Custodian will be briefed on responsibilities and expectations for the implementation of this TI. The briefing as a minimum is to cover items contained in Step 3.2.1. C and D.
 
WBN                             Containment Access                           TI-12.07A Unit 1                                Modes 1 - 4                            Rev. 0007 Page 31 of 50 Appendix A (Page 1 of 3)
Containment/Annulus Entry Authorization 1.0 AUTHORIZATION FOR GENERAL ACCESS Personnel Access To Be Used:(3)                   757 Airlock (Upper)           716 Airlock (Lower)           713 Annulus Airlock Door Alarms Enabled?                   YES                       Initials         NO(4)                   Initials Custodian Designated?                         YES                       Initials         NO(4)(6)                 Initials Incore Detectors Tagged and Stored IAW TI-41 and NO                               YES                              Initials entry inside the Polar Crane Wall while in Mode 2 or above.
Containment/Annulus Access Authorization:                       Mode(s): ____________
(1)
Entry Authorization:
Unit SRO or SM                     Date             Time (1)
Entry Authorization:
RP Supt. or designee                 Date             Time Remarks:
(1)     RP Superintendent or designee and SM/SRO signature required for all Containment entries.
(2)     Entry inside the Polar Crane Wall or Regen Hx Room while in Modes 1 & 2 REQUIRES authorization by RP Manager AND SM in accordance with RCI-128 prior to entry.
(3)     IF Upper or Lower Containment entry required with incore probes NOT tagged or NOT Stored IAW TI-41, THEN Shift Manager AND RP Manager must authorize by completion of Section 2.0, Special Access Authorization.5, 6 Requirement for Upper Containment entry based on safety concern for Containment egress if upper airlock doors become inoperable.
(4)     Signature verifies availability of Access Control Custodian for Containment, if Airlock door alarms are not enabled.
(5)     Radiological hazard NO longer present - irradiated source NO longer poses potential for personnel exposure (Incore detectors TAGGED and properly stored OR removed) AND protective measures/controls are established.
(6)     The SM may waive establishing Access Custodian in the event of urgent entry.
 
WBN                             Containment Access                           TI-12.07A Unit 1                                Modes 1 - 4                            Rev. 0007 Page 32 of 50 Appendix A (Page 2 of 3)
Containment/Annulus Entry Authorization 2.0 AUTHORIZATION FOR SPECIAL ACCESS Personnel Access To Be Used:(3)                   757 Airlock (Upper)           716 Airlock (Lower)           713 Annulus Airlock Door Alarms Enabled?                   YES                       Initials         NO(4)                   Initials Custodian Designated(3)                       YES                       Initials Authorized Work Group(s): ___________________________________________________________________
Description of Authorized Work: _______________________________________________________________
Work Limitations or Special Instructions (i.e., Access Control Requirements, specific ALARA Plan, etc.):
Incore Detectors NOT Tagged or                                       Polar Crane Wall (Modes 1 or 2)
NOT Stored IAW TI-41(3)                                 Initials     or Regen Hx Rm Entry Required(2)                           Initials Containment/Annulus Access Authorization:                           Mode(s): ______________
Entry Authorization - Incore Detectors NOT TAGGED, NOT Stored IAW TI-41 OR Crane Wall entry required while in Mode 1 or 2.(2)(3)
Shift Manager                     Date             Time Entry Authorization - Incore Detectors NOT TAGGED, NOT Stored IAW TI-41 OR Crane Wall entry required while in Mode 1 or 2.(2)(3)
RP Manager                         Date             Time (1)     RP Superintendent or designee and SM/SRO signature required for all Containment entries.
(2)     Entry inside the Polar Crane Wall or Regen Hx Room while in Modes 1 & 2 REQUIRES authorization by RP Manager AND SM in accordance with RCI-128 prior to entry.
(3)     IF Upper or Lower Containment entry required with incore probes NOT tagged or NOT Stored IAW TI-41, THEN Shift Manager AND RP Manager must authorize by completion of Section 2.0, Special Access Authorization.5, 6 Requirement for upper containment entry based on safety concern for containment egress if upper airlock doors become inoperable.
(4)     Signature verifies availability of Access Control Custodian for Containment, if Airlock door alarms are not enabled.
(5)     Radiological hazard NO longer present - irradiated source NO longer poses potential for personnel exposure (Incore detectors TAGGED and properly stored OR removed) AND protective measures/controls are established.
(6)     The SM may waive establishing Access Custodian in the event of urgent entry.
 
WBN                   Containment Access                 TI-12.07B Unit 0                      Modes 5 & 6                  Rev. 0006 Page 12 of 38 Date ________
4.1   Preliminary Actions (continued)
[8]   ENSURE personnel who will be performing inspections are briefed on the following:
* personnel safety precautions and protective measures (climbing, heat stress, ALARA, etc)
* personnel safety precautions and protective measures (climbing, heat stress, ALARA, etc)
* detailed acceptance criteria in Section 5.0 for housekeeping/cleanliness standards and expectations (Appendix E may be utilized in conducting inspections)
* detailed acceptance criteria in Section 5.0 for housekeeping/cleanliness standards and expectations (Appendix E may be utilized in conducting inspections)
* techniques for conducting an inspection, where to look and what to check, etc.
* techniques for conducting an inspection, where to look and what to check, etc.
* potential consequences of inadequate inspection on sump operability and impact to plant operation (i.e., leaking PZR Bypass Spray Valves)
* potential consequences of inadequate inspection on sump operability and impact to plant operation (i.e., leaking PZR Bypass Spray Valves)
* requirements for documenting deficiencies which cannot be immediately corrected ________ SRO [9] IF known glycol leak or RCP oil leak exists, THEN PERFORM 1-TI-12.20, Containment Formaldehyde Stay Time Calculation. ________
* requirements for documenting deficiencies which cannot be immediately corrected                                       ________
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 6 of 16  C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)] Some examples of SRO exam items for this topic include:
SRO
[9]   IF known glycol leak or RCP oil leak exists, THEN PERFORM 1-TI-12.20, Containment Formaldehyde Stay Time Calculation.                                                         ________
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]
Some examples of SRO exam items for this topic include:
* 10 CFR 50.59 screening and evaluation processes.
* 10 CFR 50.59 screening and evaluation processes.
* Administrative processes for temporary modifications.
* Administrative processes for temporary modifications.
* Administrative processes for disabling annunciators.
* Administrative processes for disabling annunciators.
* Administrative processes for the installation of temporary instrumentation.
* Administrative processes for the installation of temporary instrumentation.
* Processes for changing the plant or plant procedures. Section IV provides an example of a satisfactory SRO-only question related to this topic. D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.       [10 CFR 55.43(b)(4)] Some examples of SRO exam items for this topic include:
* Processes for changing the plant or plant procedures.
Section IV provides an example of a satisfactory SRO-only question related to this topic.
D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
[10 CFR 55.43(b)(4)]
Some examples of SRO exam items for this topic include:
Again, this question
* Process for gaseous/liquid release approvals, i.e., release permits.
* Process for gaseous/liquid release approvals, i.e., release permits.
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures. Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits. SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc. E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
is questioning the
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedure's overall mitigative strategy or purpose.  
* Analysis and interpretation of radiation and activity readings as they requirements for        pertain to selection of administrative, normal, abnormal, and emergency containment entry        procedures.
 
(an SRO function).      Analysis and interpretation of coolant activity, including comparison to The special access      emergency plan criteria and/or regulatory limits.
The applicant's knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedure's content is required to correctly answer the written test item, for example:
into containment requires the        SRO-only knowledge should not be claimed for questions that can be authorization of the answered solely based on RO knowledge of radiological safety principles; SM.                  e.g., RWP requirements, stay-time, DAC-hours, etc.
E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.
The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:
Page 6 of 16


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 7 of 16
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
* Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
* Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
SRO-only knowledge should not be claimed for questions that can be answered solely using "systems knowledge"; e.g.:
SRO-only knowledge should not be claimed for questions that can be answered solely using systems knowledge; e.g.:
* how the system works.
* how the system works.
* system flow path.
* system flow path.
* component locations, etc. SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:
* component locations, etc.
SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:
* the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
* the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
* any AOP entry condition.
* any AOP entry condition.
* plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
* plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
* immediate operator actions of a procedure. Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.         
* immediate operator actions of a procedure.
Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.
Page 7 of 16
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)
(Assessment and selection of procedures)
Can the question be answered solely by knowing systems knowledge, i.e., how the system works,      Yes RO question flowpath, logic, component location?
No Can the question be answered solely by knowing immediate operator actions?                            Yes    RO question No Can the question be answered solely by knowing entry conditions for AOPs or plant parameters          Yes RO question that require direct entry to major EOPs?
No Can the question be answered solely by knowing the purpose, overall sequence of events, or            Yes RO question overall mitigative strategy of a procedure?
No Using knowledge of the contents of Does the question require one or more of the following?                TI-12.07A, the applicant must
* Assessing plant conditions (normal, abnormal, or                  select between two emergency) and then selecting a procedure or section of a         SOI sections.
procedure to mitigate, recover, or with which to proceed
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps                                            Yes        SRO-only
* Knowledge of diagnostic steps and decision points in the              question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 8 of 16  Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5) (Assessment and selection of procedures)       
99.
Given the following timeline:
00:00:00    Unit 1 is at 100% power.
A FIRE occurs in the B trained SDBD rooms.
00:0 1:00  Unit 1 remains STABLE        at 100% power.
The following indications develop:
NOTE:        1-AOI-30.1, Plant Fires 1-AOI-30.2, Fire Safe Shutdown Which ONE of the following describes the proper implementation of the AOIs and EOPs?
A. At 00:0 1:00, 1-AOI-30.1 would continue to be in effect, because ONLY ONE train of safety related equipment is at risk.
B. At 00:00:00, 1-AOI-30.2 would be entered and would take precedence over the EOP set, SOLELY because the fire occurred in the SDBD rooms.
C. At 00:0 1:00, 1-AOI-30.2 would be entered and would NOT take precedence over the EOP set, because the Appendix R fire does not analyze for subsequent casualties.
D. At 00:0 1:00, 1-AOI-30.2 would be entered and would take precedence over the EOP set, because plant indications demonstrate that the ability of the plant to achieve and maintain safe shutdown is jeopardized.


Can the question be answered solely by knowing "systems knowledge", i.e., how the system works, flowpath, logic, component location? RO question YesNoCan the question be answered solely by knowing immediate operator actions? YesCan the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?YesNoDoes the question require one or more of the following?
CORRECT ANSWER:                                                               D DISTRACTOR ANALYSIS:
* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
A. Incorrect: At 00:00:00, 1-AOI-30.1 is in effect as there is a fire in the plant. At 00:0 1:00, conditions exist which indicate that the fire in the SDBD room has caused multiple spurious actuation (a spurious start of the 1B-B SIP and the opening of the B trained BIT valve). This indicates to the SRO that a loss of plant control is imminent and step 4 of the 1-AOI-30.1 will direct that 1-AOI-30.2 be entered to address the Appendix R fire. It is plausible to believe that 1-AOI-30.1 will continue to be in effect because one may claim that single failure criteria exists and that the loss of one train of any component does not cause a loss of safety function. One may also consider the drastic actions which are taken during an Appendix R fire and thus wait until more severe impacts are observed.
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps
B. Incorrect: 1-AOI-30.2 does state that: For an Appendix R fire, this procedure [1-AOI-30.2] takes precedence over the Emergency Operating procedures. Therefore, this portion of the distractor is correct.
* Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
However, one must note that the spurious actuations occur at 00:0 1:00. As described, 1-AOI-30.1 does not immediately transition one to 1-AOI-30.2 if a fire develops in the SDBD room. 1-AOI-30.1 contains the guidance that the SRO must observe: 1. Multiple spurious actuations of systems/components, 2. Erratic or questionable indications on numerous MCR meters/recorders or 3. Multiple trains/channels of safety related equipment involved. Then the SRO must decide that a loss of plant control is imminent and transition to 1-AOI-30.2.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?NoNoYesQuestion might not be linked to 10 CFR 55.43(b)(5) for SRO-only NoRO question RO question RO question 
This distractor is incorrect because at the very onset of the fire (at time 00:00:00) it claims that 1-AOI-30.2 is in effect because of the location of the fire. While there is a section of 1-AOI-30.2 to address a fire in the SDBD rooms, it is only used after an entry is made into 1-AOI-30.2 from 1-AOI-30.1 after the criteria aforementioned are noted.
: 99. Given the following timeline: 00:00:00 Unit 1 is at 100% power. A FIRE occurs in the "B" trained SDBD rooms. 00:0 1:00 Unit 1 remains STABLE at 100% power. The following indications develop: . Which ONE of the following describes the proper implementation of the AOIs and EOPs? A. At 00:0 1:00, 1-AOI-30.1 would continue to be in effect, because ONLY ONE train of safety related equipment is at risk. B. At 00:00:00, 1-AOI-30.2 would be entered and would take precedence over the EOP set, SOLELY because the fire occurred in the SDBD rooms. C. At 00:0 1:00, 1-AOI-30.2 would be entered and would NOT take precedence over the EOP set, because the Appendix R fire does not analyze for subsequent casualties. D. At 00:0 1:00, 1-AOI-30.2 would be entered and would take precedence over the EOP set, because plant indications demonstrate that the ability of the plant to achieve and maintain safe shutdown is jeopardized. NOTE: 1-AOI-30.1, Plant Fires  1-AOI-30.2, Fire Safe Shutdown CORRECT ANSWER:D DISTRACTOR ANALYSIS:   A. Incorrect: At 00:00:00, 1-AOI-30.1 is in effect as there is a fire in the plant. At 00:0 1:00, conditions exist which indicate that the fire in the SDBD room has caused multiple spurious actuation (a spurious start of the 1B-B SIP and the opening of the "B" trained BIT valve). This indicates to the SRO that a loss of plant control is imminent and step 4 of the 1-AOI-30.1 will direct that 1-AOI-30.2 be entered to address the Appendix R fire. It is plausible to believe that 1-AOI-30.1 will continue to be in effect because one may claim that single failure criteria exists and that the loss of one train of any component does not cause a loss of safety function. One may also consider the drastic actions which are taken during an Appendix R fire and thus wait until more severe impacts are observed. B. Incorrect: 1-AOI-30.2 does state that: "For an Appendix R fire, this procedure [1-AOI-30.2] takes precedence over the Emergency Operating procedures."  Therefore, this portion of the distractor is correct. However, one must note that the spurious actuations occur at 00:0 1:00. As described, 1-AOI-30.1 does not immediately transition one to 1-AOI-30.2 if a fire develops in the SDBD room. 1-AOI-30.1 contains the guidance that the SRO must observe: "1. Multiple spurious actuations of systems/components, 2. Erratic or questionable indications on numerous MCR meters/recorders or 3. Multiple trains/channels of safety related equipment involved."  Then the SRO must decide that a loss of plant control is imminent and transition to 1-AOI-30.2.
C. Incorrect: The background seen in 1-AOI-30.2 (pg 12) relates: I. No other accident is assumed to occur concurrently with a fire. Therefore, while it is true that the Appendix R fire does not analyze for subsequent casualties, as mentioned previously it is not true that the EOP set takes precedence over 1-AOI-30.2.
This distractor is incorrect because at the very onset of the fire (at time 00:00:00) it claims that 1-AOI-30.2 is in effect because of the location of the fire. While there is a section of 1-AOI-30.2 to address a fire in the SDBD rooms, it is only used after an entry is made into 1-AOI-30.2 from 1-AOI-30.1 after the criteria aforementioned are noted. C. Incorrect: The background seen in 1-AOI-30.2 (pg 12) relates: "I. No other accident is assumed to occur concurrently with a fire."  Therefore, while it is true that the Appendix R fire does not analyze for subsequent casualties, as mentioned previously it is not true that the EOP set takes precedence over 1-AOI-30.2.
D. Correct: At 00:0 1:00, indications exist that support an entry into 1-AOI-30.2. As seen in step 3 of 1-AOI-30.1, the fire has demonstrated the potential to affect plant control (safe shutdown capability). As already mentioned, 1-AOI-30.2 takes precedence over the EOP set.
D. Correct: At 00:0 1:00, indications exist that support an entry into 1-AOI-30.2. As seen in step 3 of 1-AOI-30.1, the fire has demonstrated the "potential to affect plant control (safe shutdown capability)."  As already mentioned, 1-AOI-30.2 takes precedence over the EOP set.
Question Number: 99  Tier:  3 Group:
K/A: G 2.4.23 2.4 Emergency Procedures / Plan Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations.
Importance Rating: 3.6  4.4


10 CFR Part 55: (CFR: 41.7 / 41.10 / 43.5 / 45.12) 10CFR55.43.b: 10 CFR 55.43(b)(5) K/A Match: K/A is matched because the applicant is required to understand the basis behind the preferential implementation of the appropriate procedure set (either the emergency or abnormal operating procedures) for various casualties. Technical  
Question Number:      99 Tier:    3  Group:
K/A:    G 2.4.23 2.4 Emergency Procedures / Plan Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations.
Importance Rating:    3.6 4.4 10 CFR Part 55:     (CFR: 41.7 / 41.10 / 43.5 / 45.12) 10CFR55.43.b:       10 CFR 55.43(b)(5)
K/A Match:   K/A is matched because the applicant is required to understand the basis behind the preferential implementation of the appropriate procedure set (either the emergency or abnormal operating procedures) for various casualties.
Technical  


==Reference:==
==Reference:==
1-AOI-30.1, Plant Fires 1-AOI-30.2, Fire Safe Shutdown Proposed references to be provided: None  Learning Objective: 3-OT-AOI-3000, Plant Fires 4. Given a set of plant conditions, DESCRIBE operator actions required in accordance with 1-AOI-30. Cognitive Level:     Higher X Lower     Question Source:     New X Modified Bank   Bank     Question History: New question for the 2015-301 NRC SRO Exam Comments: The question is SRO only as detailed in the distractor analysis. The question meets the general SRO only criteria of "Assessment of facility conditions an selection of appropriate procedures during normal, abnormal, and emergency situations."
1-AOI-30.1, Plant Fires 1-AOI-30.2, Fire Safe Shutdown Proposed references to   None be provided:
WBN Unit 1 Plant Fires 1-AOI-30.1 Rev. 0002    Page 4 of 11  Step Action/Expected Response Response Not Obtained 3.0 OPERATOR ACTIONS 1. IF a valid verbal report, annunciation or indication of a fire is present, THEN REQUEST UO perform Appendix B. 2. ENSURE both trains Control Room Isolation (CRI) signals are DARK on Master Status Panel (MSP) on M-6. REFER TO SOI-31.01, Control Building HVAC System, to evaluate aligning CREVS suction to the other side of the Control Building. NOTE The decision to trip the Unit and declare an Appendix R fire is left to the judgment of the Unit SRO/Shift Manager and must be based on the magnitude of the fire and its potential effect on the equipment/components necessary to achieve and maintain cold shutdown. 3. MONITOR magnitude of the fire and the potential to affect plant control (safe shutdown capability):
Learning Objective:       3-OT-AOI-3000, Plant Fires
* Multiple spurious actuations of systems/components.
: 4. Given a set of plant conditions, DESCRIBE operator actions required in accordance with 1-AOI-30.
* Erratic or questionable indications on numerous MCR meters/recorders.
Cognitive Level:
* Multiple trains/channels of safety related equipment involved. 4. IF loss of plant control is imminent or becomes imminent during the performance of this Instruction, THEN
Higher             X Lower Question Source:
** GO TO AOI-30.2, Fire Safe Shutdown.
New               X Modified Bank Bank Question History:         New question for the 2015-301 NRC SRO Exam Comments:                 The question is SRO only as detailed in the distractor analysis. The question meets the general SRO only criteria of Assessment of facility conditions an selection of appropriate procedures during normal, abnormal, and emergency situations.
WBN Unit 1 Fire Safe Shutdown 1-AOI-30.2 Rev. 0004  Step    Action/Expected Response   Response Not Obtained   Page 4 of 22  4.0 OPERATOR ACTIONS   NOTES The decision to trip the unit and declare an Appendix R fire is left to the judgment of the Unit SRO/SM and must be based on the magnitude of the fire and its potential effect on the equipment and components necessary to achieve and maintain cold shutdown For an Appendix R fire, this procedure takes precedence over the Emergency Operating Procedures AUO local operator actions should be assigned as early as possible by an SRO or UO NOT involved with immediate actions of this procedure. 1. DETERMINE the fire location has the potential to affect equipment needed for safe shutdown. RETURN to 1-AOI 30.1  NOTE For a fire that touches a soft interface (NO physical wall or barrier), as indicated by heavy dashed lines in 1-AOI-30.2 APP B, choose the room where the fire is predominate. When the fire is basically centered between the rooms the actions of either room are sufficient. 2. REFER to 1-AOI-30.2 APP B, Elevation Diagrams, to determine applicable 1-AOI-30.2 C-Series appendix.
 
WBN Unit 1 Fire Safe Shutdown 1-AOI-30.2 Rev. 0004 5.1 Background and Assumptions (continued)   Page 12 of 22  G. The Safe Shutdown Logic Diagram also shows the paths available to provide the safety functions for the safe shutdown conditions described in Paragraphs 5.1E and 5.1F. For each safety function, the equipment required to accomplish the safety function has been divided into "Keys" which represent groups of functionally-related equipment necessary to accomplish the safety function. These are also represented on the Safe Shutdown Logic Diagram. H. At least one path of equipment or components needed to achieve safe shutdown is required to remain operable or capable of being operated for 72 hours following a postulated fire (to establish long-term heat removal via RHR). I. No other accident is assumed to occur concurrently with a fire therefore, a valid SI signal is assumed not to be present at the time of an Appendix R fire. However, spurious SI signal actuation could occur as a result of the effects of the fire. Since many of the actions in the Safe Shutdown analysis require components to be in positions opposite that required by SI, a spurious SI would require these components to be repositioned.
WBN                           Plant Fires                 1-AOI-30.1 Unit 1                                                    Rev. 0002 Step   Action/Expected Response                     Response Not Obtained 3.0     OPERATOR ACTIONS
For example, the BIT outlet valves are required to be closed for an Appendix R fire. The purpose of this is to: 1. Guarantee flow to the RCP seal line for boron injection 2. Prevent pressurizer overfill (no RCS break is assumed and normal charging/letdown may not be available due to fire or loss of air). 3. Prevent damage to the charging pump due to fast drawdown of the VCT (automatic circuit for the swap over to RWST on low VCT level is not guaranteed). J. In general it is assumed that shutdown of the plant will be performed from the Main Control Room for a postulated fire elsewhere in the plant. For shutdown from outside the Control Building, it is essential that, functionally, the same equipment and instrumentation be available from the Aux Control Room or remote stations or otherwise be justified. Loss of offsite power is assumed concurrently with MCR fires. K. The possibility of shutdown and cooldown of the plant from the auxiliary control room was considered in the manual actions of the approved Site Engineering calculation. L. Where the spurious operation of a component could defeat the required system safety function, manual actions are taken to address the effects of spurious component operation. Components identified as those which could prevent a safe shutdown should they spurious operate are those that:
: 1.     IF a valid verbal report, annunciation or indication of a fire is present, THEN REQUEST UO perform Appendix B.
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 6 of 16  C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)] Some examples of SRO exam items for this topic include:
: 2.     ENSURE both trains Control Room             REFER TO SOI-31.01, Control Building Isolation (CRI) signals are DARK on         HVAC System, to evaluate aligning Master Status Panel (MSP) on M-6.           CREVS suction to the other side of the Control Building.
NOTE           The decision to trip the Unit and declare an Appendix R fire is left to the judgment of the Unit SRO/Shift Manager and must be based on the magnitude of the fire and its potential effect on the equipment/components necessary to achieve and maintain cold shutdown.
: 3.     MONITOR magnitude of the fire and                                 The concern is the fire's the potential to affect plant control                             ability to affect the (safe shutdown capability):                                       capability of the plant to achieve and maintain a
* Multiple spurious actuations of safe shutdown.
systems/components.
* Erratic or questionable indications on numerous MCR meters/recorders.                                       These are criteria used by
* Multiple trains/channels of safety                     the SRO to determine if an related equipment involved.                             Appendix R fire is to be declared (i.e. if a transition
: 4.     IF loss of plant control is imminent or                     to 1-AOI-30.2 is warranted).
becomes imminent during the performance of this Instruction, THEN The stem of the
        ** GO TO AOI-30.2, Fire Safe                             question shows Shutdown.                                                that a loss of plant control is imminent.
Page 4 of 11
 
WBN                     Fire Safe Shutdown                 1-AOI-30.2 Unit 1                                                      Rev. 0004 Step   Action/Expected Response                     Response Not Obtained 4.0     OPERATOR ACTIONS NOTES
* The decision to trip the unit and declare an Appendix R fire is left to the judgment of the Unit SRO/SM and must be based on the 1-AOI-30.2 trumps              magnitude of the fire and its potential effect on the equipment and the EOP set.                    components necessary to achieve and maintain cold shutdown
* For an Appendix R fire, this procedure takes precedence over the Emergency Operating Procedures
* AUO local operator actions should be assigned as early as possible by an SRO or UO NOT involved with immediate actions of this procedure.
: 1.       DETERMINE the fire location has           RETURN to 1-AOI 30.1 the potential to affect equipment needed for safe shutdown.
NOTE       For a fire that touches a soft interface (NO physical wall or barrier), as indicated by heavy dashed lines in 1-AOI-30.2 APP B, choose the room where the fire is predominate. When the fire is basically centered between the rooms the actions of either room are sufficient.
: 2.       REFER to 1-AOI-30.2 APP B,                                   Notice that while the SDBD Elevation Diagrams, to determine rooms have a C-series applicable 1-AOI-30.2 C-Series appendix associated with them, appendix.
a fire within such rooms does not automatically cause the use of this procedure.
Page 4 of 22
 
WBN                       Fire Safe Shutdown               1-AOI-30.2 Unit 1                                                      Rev. 0004 5.1     Background and Assumptions (continued)
G. The Safe Shutdown Logic Diagram also shows the paths available to provide the safety functions for the safe shutdown conditions described in Paragraphs 5.1E and 5.1F. For each safety function, the equipment required to accomplish the safety function has been divided into Keys which represent groups of functionally-related equipment necessary to accomplish the safety function. These are also represented on the Safe Shutdown Logic Diagram.
H. At least one path of equipment or components needed to achieve safe shutdown is required to remain operable or capable of being operated for 72 hours following a postulated fire (to establish long-term heat removal via RHR).
I. No other accident is assumed to occur concurrently with a fire therefore, a valid SI signal is assumed not to be present at the time of an Appendix R fire.
However, spurious SI signal actuation could occur as a result of the effects of the fire. Since many of the actions in the Safe Shutdown analysis require No other accident    components to be in positions opposite that required by SI, a spurious SI would is assumed to        require these components to be repositioned.
occur with the Appendix R fire.      For example, the BIT outlet valves are required to be closed for an Appendix R fire. The purpose of this is to:
: 1. Guarantee flow to the RCP seal line for boron injection
: 2. Prevent pressurizer overfill (no RCS break is assumed and normal charging/letdown may not be available due to fire or loss of air).
: 3. Prevent damage to the charging pump due to fast drawdown of the VCT (automatic circuit for the swap over to RWST on low VCT level is not guaranteed).
J. In general it is assumed that shutdown of the plant will be performed from the Main Control Room for a postulated fire elsewhere in the plant. For shutdown from outside the Control Building, it is essential that, functionally, the same equipment and instrumentation be available from the Aux Control Room or remote stations or otherwise be justified. Loss of offsite power is assumed concurrently with MCR fires.
K. The possibility of shutdown and cooldown of the plant from the auxiliary control room was considered in the manual actions of the approved Site Engineering calculation.
L. Where the spurious operation of a component could defeat the required system safety function, manual actions are taken to address the effects of spurious component operation. Components identified as those which could prevent a safe shutdown should they spurious operate are those that:
Page 12 of 22
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]
Some examples of SRO exam items for this topic include:
* 10 CFR 50.59 screening and evaluation processes.
* 10 CFR 50.59 screening and evaluation processes.
* Administrative processes for temporary modifications.
* Administrative processes for temporary modifications.
* Administrative processes for disabling annunciators.
* Administrative processes for disabling annunciators.
* Administrative processes for the installation of temporary instrumentation.
* Administrative processes for the installation of temporary instrumentation.
* Processes for changing the plant or plant procedures. Section IV provides an example of a satisfactory SRO-only question related to this topic. D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.       [10 CFR 55.43(b)(4)] Some examples of SRO exam items for this topic include:
* Processes for changing the plant or plant procedures.
Section IV provides an example of a satisfactory SRO-only question related to this topic.
D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
[10 CFR 55.43(b)(4)]
Some examples of SRO exam items for this topic include:
* Process for gaseous/liquid release approvals, i.e., release permits.
* Process for gaseous/liquid release approvals, i.e., release permits.
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures. Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits. SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc. E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedure's overall mitigative strategy or purpose.  
Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.
SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.
E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.
The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:
Page 6 of 16


The applicant's knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedure's content is required to correctly answer the written test item, for example: 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 7 of 16
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
* Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
* Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
SRO-only knowledge should not be claimed for questions that can be answered solely using "systems knowledge"; e.g.:
SRO-only knowledge should not be claimed for questions that can be answered solely using systems knowledge; e.g.:
* how the system works.
* how the system works.
* system flow path.
* system flow path.
* component locations, etc. SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:
* component locations, etc.
SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:
* the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
* the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
* any AOP entry condition.
* any AOP entry condition.
* plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
* plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
* immediate operator actions of a procedure. Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.        
* immediate operator actions of a procedure.
Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.
Page 7 of 16


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 8 of 16  Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5) (Assessment and selection of procedures)      
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)
(Assessment and selection of procedures)
Can the question be answered solely by knowing systems knowledge, i.e., how the system works,      Yes RO question flowpath, logic, component location?                              Note that this question does not No question the entry condition for the 1-Can the question be answered solely by knowing immediate operator actions?                            Yes    RO question (as such AOI-30.2 procedure is No                                                    entered by transitioning from 1-AOI-30.1).
Can the question be answered solely by knowing entry conditions for AOPs or plant parameters          Yes RO question that require direct entry to major EOPs?
No Can the question be answered solely by knowing                      This question requires the purpose, overall sequence of events, or            Yes RO question that the applicant overall mitigative strategy of a procedure?
understand the transition No                                                      from 1-AOI-30.1 to 1-AOI-30.2 AND the Does the question require one or more of the following?            prioritization of the appendix R fire recovery
* Assessing plant conditions (normal, abnormal, or              with respect to the EOP emergency) and then selecting a procedure or section of a      set.
procedure to mitigate, recover, or with which to proceed
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps                                            Yes          SRO-only
* Knowledge of diagnostic steps and decision points in the                question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16


Can the question be answered solely by knowing "systems knowledge", i.e., how the system works, flowpath, logic, component location? RO question YesNoCan the question be answered solely by knowing immediate operator actions? YesCan the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?YesNoDoes the question require one or more of the following?
100.
* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
Which ONE of the following describes the Protective Action Recommendations (PARs)?
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps
* Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?NoNoYesQuestion might not be linked to 10 CFR 55.43(b)(5) for SRO-only NoRO question RO question RO question 100. Which ONE of the following describes the Protective Action Recommendations (PARs)?
In accordance with EPIP-4, Site Area Emergency, PARs are ____(1)____.
In accordance with EPIP-4, Site Area Emergency, PARs are ____(1)____.
In accordance with EPIP-1, Emergency Plan Classification Logic, ____(2)____ can assume the responsibility for PARs when the respective emergency center is staffed and operational.
In accordance with EPIP-1, Emergency Plan Classification Logic, ____(2)____
A. (1) optional (2) CECC director B. (1) optional (2) TSC RP Manager C. (1) NOT made (2) CECC director D. (1) NOT made (2) TSC RP Manager CORRECT ANSWER:C DISTRACTOR ANALYSIS:   A. Incorrect: IAW REP Generic and EPIP-5 GE; PARs are only made at the General Emergency Level. REP Generic and EPIP-4, SAE do not direct a PAR be made. IAW EPIP-1 and CECC EPIP-1, the CECC Director is the ONLY person that may assume the responsibility for PARs. B. Incorrect: IAW REP Generic and EPIP-5 GE; PARs are only made at the General Emergency Level. REP Generic and EPIP-4, SAE do not direct a PAR be made. There are responsibilities that the CECC Director may delegate; this is not one of those responsibilities. The TSC RP Manger makes recommendations to the SED and CECC Director on Protective requirements and dose management. C. Correct: IAW REP Generic and EPIP-5 GE; PARs are only made at the General Emergency Level. REP Generic and EPIP-4, SAE do not direct a PAR be made. There are responsibilities that the CECC Director may delegate; this is not one of those responsibilities. D. Incorrect: IAW REP Generic and EPIP-5 GE; PARs are only made at the General Emergency Level. REP Generic and EPIP-4, SAE do not direct a PAR be made. There are responsibilities that the CECC Director may delegate; this is not one of those responsibilities. The TSC RP Manger makes recommendations to the SED and CECC Director on Protective requirements and dose management.
can assume the responsibility for PARs when the respective emergency center is staffed and operational.
Question Number: 100 Tier:  3 Group:
A.   (1)   optional (2)   CECC director B.   (1)   optional (2)   TSC RP Manager C.   (1)   NOT made (2)   CECC director D.   (1)   NOT made (2)   TSC RP Manager
K/A: 2.4 Emergency Procedures / Plan 2.4.44 Knowledge of emergency plan protective action recommendations. Importance Rating: 2.4  4.4
 
CORRECT ANSWER:                                                         C DISTRACTOR ANALYSIS:
A. Incorrect: IAW REP Generic and EPIP-5 GE; PARs are only made at the General Emergency Level. REP Generic and EPIP-4, SAE do not direct a PAR be made.
IAW EPIP-1 and CECC EPIP-1, the CECC Director is the ONLY person that may assume the responsibility for PARs.
B. Incorrect: IAW REP Generic and EPIP-5 GE; PARs are only made at the General Emergency Level. REP Generic and EPIP-4, SAE do not direct a PAR be made.
There are responsibilities that the CECC Director may delegate; this is not one of those responsibilities.
The TSC RP Manger makes recommendations to the SED and CECC Director on Protective requirements and dose management.
C. Correct: IAW REP Generic and EPIP-5 GE; PARs are only made at the General Emergency Level. REP Generic and EPIP-4, SAE do not direct a PAR be made.
There are responsibilities that the CECC Director may delegate; this is not one of those responsibilities.
D. Incorrect: IAW REP Generic and EPIP-5 GE; PARs are only made at the General Emergency Level. REP Generic and EPIP-4, SAE do not direct a PAR be made.
There are responsibilities that the CECC Director may delegate; this is not one of those responsibilities.
The TSC RP Manger makes recommendations to the SED and CECC Director on Protective requirements and dose management.


10 CFR Part 55: (CFR: 41.10 / 41.12 / 43.5 / 45.11) 10CFR55.43.b: 10 CFR 55.43(b)(5)
Question Number:      100 Tier:    3  Group:
K/A Match: KA is matched because the question requires knowledge of emergency plan protective action recommendations.
K/A:    2.4 Emergency Procedures / Plan 2.4.44 Knowledge of emergency plan protective action recommendations.
Importance Rating:    2.4 4.4 10 CFR Part 55:     (CFR: 41.10 / 41.12 / 43.5 / 45.11) 10CFR55.43.b:       10 CFR 55.43(b)(5)
K/A Match:   KA is matched because the question requires knowledge of emergency plan protective action recommendations.
Technical  
Technical  


==Reference:==
==Reference:==
REP Generic EPIP-5, GE EPIP-4, SAE EPIP-1, Emergency Plan Classification Logic CECC EPIP-1 Proposed references to be provided: None  Learning Objective: 3-OT-PCD-048C, Radiological Emergency Plan 8. Given a plant situation that requires PARs, determine the correct PARs IAW EPIP-5. Cognitive Level:     Higher   Lower X   Question Source:     New   Modified Bank   Bank X   Question History: Bank question 2.4.44 700, last used on the WBN 08/2010 NRC exam. Comments:
REP Generic EPIP-5, GE EPIP-4, SAE EPIP-1, Emergency Plan Classification Logic CECC EPIP-1 Proposed references to   None be provided:
Radiological Emergency Procedure RADIOLOGICAL EMERGENCY PLAN (GENERIC PART) REP-Generic Rev. 0104 Page 33 of 90 5.2.2 Alert (continued)     1. Class of emergency. 2. Type of actual or projected release (airborne, waterborne, or surface spill) and estimated duration/impact times. 3. Estimate of quantity of radioactive material released or being released and the height of release. 4. Chemical and physical form of released material, including estimates of the relative quantities and concentration of noble gases, iodines, and particulates. 5. Prevailing weather (wind velocity, direction, temperature, atmospheric stability data, and form of precipitation, if any). 6. Actual or projected doses at site boundary. 7. Projected dose rates and integrated dose at about 2, 5, and 10 miles, including sector(s) affected. 8. Estimate of any surface spill radioactive contamination.  
Learning Objective:       3-OT-PCD-048C, Radiological Emergency Plan
: 9. Emergency response actions underway. 10. Request for any needed onsite support by offsite organizations. 11. Prognosis for worsening or termination of event based on plant information. G. The JIC may be activated.
: 8. Given a plant situation that requires PARs, determine the correct PARs IAW EPIP-5.
Cognitive Level:
Higher Lower             X Question Source:
New Modified Bank Bank               X Question History:         Bank question 2.4.44 700, last used on the WBN 08/2010 NRC exam.
Comments:
 
Radiological         RADIOLOGICAL EMERGENCY PLAN                   REP-Generic Emergency                        (GENERIC PART)                     Rev. 0104 Procedure                                                          Page 33 of 90 5.2.2   Alert (continued)
: 1. Class of emergency.
: 2. Type of actual or projected release (airborne, waterborne, or surface spill) and estimated duration/impact times.
: 3. Estimate of quantity of radioactive material released or being released and the height of release.
: 4. Chemical and physical form of released material, including estimates of the relative quantities and concentration of noble gases, iodines, and particulates.
: 5. Prevailing weather (wind velocity, direction, temperature, atmospheric stability data, and form of precipitation, if any).
: 6. Actual or projected doses at site boundary.
: 7. Projected dose rates and integrated dose at about 2, 5, and 10 miles, including sector(s) affected.
: 8. Estimate of any surface spill radioactive contamination.
: 9. Emergency response actions underway.
: 10. Request for any needed onsite support by offsite organizations.
: 11. Prognosis for worsening or termination of event based on plant information.
G. The JIC may be activated.
H. Periodic media releases are provided.
H. Periodic media releases are provided.
I. The SED augments plant shift personnel, as necessary, to initiate corrective and protective actions. 5.2.3 Site Area Emergency Upon declaration of this class: A. All the actions performed in section 5.2.2 are performed.
I. The SED augments plant shift personnel, as necessary, to initiate corrective and protective actions.
B. Personnel knowledgeable of plant systems are dispatched to the SEOC. Upon notification, these individuals should arrive at the applicable emergency operations center within a timeframe limited only by their commuting time. C. Any appropriate protective actions for the public are recommended to State agencies by the CECC. D. The JIC is activated.
5.2.3   Site Area Emergency Upon declaration of this class:
Radiological Emergency Procedure RADIOLOGICAL EMERGENCY PLAN (GENERIC PART) REP-Generic Rev. 0104 Page 34 of 90   5.2.4 General Emergency Upon declaration of this class: A. All the actions performed in section 5.2.3 are performed.
A. All the actions performed in section 5.2.2 are performed.
B. Appropriate protective action recommendations to the State are required upon declaration of General Emergency. C. If this is the initial classification, the MCR notifies the local government agencies within 15 minutes, and passes along the protective action recommendations. 5.3 Transportation Accidents 5.3.1 Notification by Carrier In the event of a transportation accident involving a TVA shipment of radioactive materials, the carrier (or other person at the accident site) contacts the ODS. The carrier has procedures outlining the notifications. 5.3.2 Notification of ODS A. State B. EDO C. Shift Manager of the Affected Site D. CECC Director E. Radiological Assessment Manager 5.3.3 CECC Director Actions The CECC Director notifies the NRC, DOT, State authorities, ANI, and DOE (information only). The appropriate State agency, NRC, ANI, and DOE have duty officers available 24 hours a day to facilitate notification of their respective agencies. 5.3.4 Radiological Assessment Manager Actions The Radiological Assessment Manager will dispatch a radiological monitoring team, if deemed necessary by the CECC Director or requested by the appropriate State agency. A Radioactive Material Specialist may be sent with the team. The TVA Representative at the scene will be the senior TVA person at the site of the incident.
B. Personnel knowledgeable of plant systems are dispatched to the SEOC. Upon notification, these individuals should arrive at the applicable emergency operations center within a timeframe limited only by their commuting time.
WBN Unit 0 General Emergency EPIP-5 Rev. 0043 Page 12 of 34   3.7 Notification of the Nuclear Regulatory Commission (NRC)   CAUTION Notification of the NRC shall be completed immediately after notification of the appropriate State or local agencies and not later than one hour after the time of Emergency Classification. [1] COMPLETE Appendix D, "Notification of the NRC". NOTE When the TSC is staffed, the open and continuous line of communications with the NRC may be transferred to the NRC Coordinator position. [2] IF REQUESTED by the NRC, DIRECT a member of the Operations staff (SRO if Available) to maintain an open and continuous line of communications as directed by NRC. 3.8 Monitor / Re-evaluate the Event [1] Monitoring and reevaluation of plant events along with communicating significant changes should be performed continuously as a function of the emergency response. Methods used to communicate significant changes are not formalized and may vary depending upon staffing levels as well as availability of personnel or equipment. Appendix E provides a systematic approach to monitoring/re-evaluation and the communication of significant changes in plant conditions. CONTINUE to conduct State follow-ups until the CECC has assumed State communications responsibilities using Appendix F "General Emergency Follow-up Information Form" to communicate follow-up information.   [2] Reevaluation of significant changes must additionally include a determination of whether Protective Action Recommendations (PARs) should be upgraded. The need to upgrade PARs is determined through the continuous assessment of Appendix H "Initial Protective Action Recommendations". IF it has been determined that a PAR Upgrade is appropriate, THEN COMMUNICATE the Upgrade to the State using Appendix J "Upgrade-Protective Action Recommendation".   [3] CONTINUE to assess PARs until the CECC has assumed PAR responsibilities.
C. Any appropriate protective actions for the public are recommended to State agencies by the CECC.
WBN Unit 0 General Emergency EPIP-5 Rev. 0043 Page 15 of 34   Appendix A (Page 1 of 1) General Emergency Initial Notification Form 1. This is a Drill This is an Actual Event - Repeat - This is an Actual Event 2. The SED at Watts Bar has declared a GENERAL EMERGENCY 3. EAL Designator:_______________________(Use only one EAL) 4. Radiological Conditions: (Check one under both Airborne and Liquid column.) Airborne Releases Offsite Liquid Releases Offsite   Minor releases within federally approved limits1 Releases above federally approved limits1   Release information not known   (1Tech Specs/ODCM)   Minor releases within federally approved limits1  Releases above federally approved limits1  Release information not known   (1Tech Specs/ODCM) 5. Event Declared: Time:__________________ Date:_____________________ 6. The Meteorological Conditions are: (Use 46 meter data from the Met Tower. IF data is NOT available from the MET tower, contact the National Weather Service by dialing 9-1-423-586-8400. The National Weather Service will provide wind direction and wind speed.) Wind Direction is FROM:_____________degrees                     (15 minute average) Wind Speed: ______________m.p.h     (15 minute average)
D. The JIC is activated.
: 7. Provide Protective Action Recommendation utilizing Appendix H: (Check either 1 or 2 or 3) Recommendation 1 RECOMENDATION 1 WIND FROM DEGREES (Mark wind direction from step 6) RECOMENDATION 2 Recommendation 2  EVACUATE LISTED SECTORS (2 mile Radius and 10 miles downwind) EVACUATE LISTED SECTORS(2 mile Radius and 5 miles downwind) SHELTER remainder of 10 mile EPZ SHELTER remainder of 10 mile EPZ CONSIDER issuance of POTASSIUM IODIDE in accordance with the State Plan CONSIDER issuance of POTASSIUM IODIDE in accordance with the State Plan A1, B1, C1, D1, C7, C9, D2, D4, D5, D6, D7, D8, D9 From 26-68 A1, B1, C1, D1, C7, D2, D4, D5 A1, B1, C1, D1, A3, A4, D2, D3, D4, D5, D6, D7, D8, D9 From 69-110 A1, B1, C1, D1, A3, D2, D4, D5 A1, B1, C1, D1, A2, A3, A4, A5, A6, A7, D2, D3, D5, D6 From 111-170 A1, B1, C1, D1, A2, A3, D2, D5 A1, B1, C1, D1, A2, A3, A5, A6, A7, B2, B3, B4, B5, C2 From 171-230 A1, B1, C1, D1, A2, A3, B2, B4, C2 A1, B1, C1, D1, B2, B3, B4, B5, C2, C3, From 231-270 A1, B1, C1, D1, B2, B4, C2 A1, B1, C1, D1, B2, B3, C2, C3, C4, C5, C6, C11 From 271-325 A1, B1, C1, D1, B2, C2, C4, C5, A1, B1, C1, D1, C2, C4, C5, C6, C7, C8, C9, C10, C11, D4, D9 From 326-25 A1, B1, C1, D1, C2, C4, C5, C7, C8, D4 Recommendation 3 SHELTER all sectors CONSIDER issuance of POTASSIUM IODIDE in accordance with the State Plan Completed by (SED)______________ Peer Checked by ________________
 
WBN Unit 0 General Emergency EPIP-5 Rev. 0043 Page 30 of 34   Appendix H (Page 1 of 2) Initial - Protective Action Recommendations WBN Unit 0 General Emergency EPIP-5 Rev. 0043 Page 31 of 34 Appendix H (Page 2 of 2) Initial - Protective Action Recommendations WBN Unit 0 Site Area Emergency EPIP-4 Rev. 0038 Page 15 of 28   Appendix A (Page 1 of 1) Site Area Emergency Initial Notification Form   1. This is a Drill   This is an Actual Event - Repeat - This is an Actual Event 2. The SED at Watts Bar has declared a Site Area Emergency 3. EAL Designator: ________________________________ 4. Radiological Conditions: (Check one under both Airborne and Liquid column.) Airborne Releases Offsite Liquid Releases Offsite   Minor releases within federally approved limits1    Minor releases within federally approved limits1   Releases above federally approved limits1   Releases above federally approved limits1   Release information not known       (1Tech Specs/ODCM)   Release information not known      (1Tech Specs/ODCM) 5. Event Declared: Time:________________ Eastern Time Date:__________________ 6. Provide Protective Action Recommendation:   None Completed By (SED)_______________________________________ Peer Checked By: _______________________________________
Radiological           RADIOLOGICAL EMERGENCY PLAN                       REP-Generic Emergency                            (GENERIC PART)                       Rev. 0104 Procedure                                                                Page 34 of 90 5.2.4   General Emergency Upon declaration of this class:
WBN Unit 0 Emergency Plan Classification Logic EPIP-1 Rev. 0042 Page 8 of 54   1.0 PURPOSE This Procedure provides guidance in determining the classification and declaration of an emergency based on plant conditions. 2.0 RESPONSIBILITY The responsibility of declaring an Emergency based on the guidance within this procedure belongs to the Shift Manager/Site Emergency Director (SM/SED) or designated Unit Supervisor (US) when acting as the SM or the TSC Site Emergency Director (SED). The following duties CAN NOT be delegated: Emergency Classification, Emergency Dose Approval and PAR development prior to CECC Director ownership for PAR development. 3.0 INSTRUCTIONS 3.1 Precautions and Limitations CAUTION Unit-2 radiation monitor readings for classification purposes do not apply until Unit-2 is licensed and operating. A. The criteria in WBN EPIP-1 are given for GUIDANCE ONLY: knowledge of actual plant conditions or the extent of the emergency may require that additional steps be taken. In all cases, this logic procedure should be combined with the sound judgment of the SM/SED and/or the TSC SED to arrive at a classification for a particular set of circumstances. B. The Nuclear Power (NP) Radiological Emergency Plan (REP) will be activated when any one of the conditions listed in this logic is detected. C. The SM/SED shall assess, classify, and declare an emergency condition within 15 minutes after information is first available to plant operators to recognize that an EAL has been exceeded and to make the declaration promptly upon identification of the appropriate Emergency Classification Level (ECL). 1. For EAL thresholds that specify duration of the off-normal condition, the emergency declaration process runs concurrently with the specified threshold duration. a. Consider as an example, the EAL "fire which is not extinguished within 15 minutes of detection." On receipt of a fire alarm, the plant fire brigade is dispatched to the scene to begin fire suppression efforts.
A. All the actions performed in section 5.2.3 are performed.
CENTRAL EMERGENCY CONTROL CENTER (CECC) OPERATIONS CECC EPIP-1 Page 6 of 44 Revision 60   3.3 CECC Director/Assistant CECC Director The CECC Director is responsible for directing TVA's overall response to the emergency. An Assistant CECC Director (who is qualified as a CECC Director) may be used to assist the CECC Director in the accomplishment of position duties. The CECC Director, at his discretion, may delegate the accomplishment of duties to the Assistant CECC Director including signature authority.
B. Appropriate protective action recommendations to the State are required upon declaration of General Emergency.
The CECC Director ensures that Federal, State, and local agencies are notified in accordance with established procedures and that they are kept fully informed of all aspects of the emergency. The Director reviews with the Plant Assessment and Radiological Assessment Managers the onsite and offsite consequences of the accident and assesses the adequacy and need for measures taken for protection of the public. The Director coordinates TVA's efforts with State and Federal agencies involved in the offsite aspects of the emergency and requests any required federal assistance through the NRC. Checklists for the CECC Director are provided in Appendices B through G. The CECC Director shall complete Appendix B for initial activation and the appropriate Appendix for the event (e.g., Appendix E for an Alert). After the appropriate level of CECC activation the CECC Director is responsible for the following:
C. If this is the initial classification, the MCR notifies the local government agencies within 15 minutes, and passes along the protective action recommendations.
5.3     Transportation Accidents 5.3.1   Notification by Carrier In the event of a transportation accident involving a TVA shipment of radioactive materials, the carrier (or other person at the accident site) contacts the ODS. The carrier has procedures outlining the notifications.
5.3.2   Notification of ODS A. State B. EDO C. Shift Manager of the Affected Site D. CECC Director E. Radiological Assessment Manager 5.3.3   CECC Director Actions The CECC Director notifies the NRC, DOT, State authorities, ANI, and DOE (information only). The appropriate State agency, NRC, ANI, and DOE have duty officers available 24 hours a day to facilitate notification of their respective agencies.
5.3.4   Radiological Assessment Manager Actions The Radiological Assessment Manager will dispatch a radiological monitoring team, if deemed necessary by the CECC Director or requested by the appropriate State agency. A Radioactive Material Specialist may be sent with the team. The TVA Representative at the scene will be the senior TVA person at the site of the incident.
 
WBN                       General Emergency             EPIP-5 Unit 0                                                    Rev. 0043 Page 12 of 34 3.7     Notification of the Nuclear Regulatory Commission (NRC)
CAUTION Notification of the NRC shall be completed immediately after notification of the appropriate State or local agencies and not later than one hour after the time of Emergency Classification.
[1]   COMPLETE Appendix D, Notification of the NRC.
NOTE When the TSC is staffed, the open and continuous line of communications with the NRC may be transferred to the NRC Coordinator position.
[2]   IF REQUESTED by the NRC, DIRECT a member of the Operations staff (SRO if Available) to maintain an open and continuous line of communications as directed by NRC.
3.8     Monitor / Re-evaluate the Event
[1]   Monitoring and reevaluation of plant events along with communicating significant changes should be performed continuously as a function of the emergency response. Methods used to communicate significant changes are not formalized and may vary depending upon staffing levels as well as availability of personnel or equipment.
Appendix E provides a systematic approach to monitoring/re-evaluation and the communication of significant changes in plant conditions.
CONTINUE to conduct State follow-ups until the CECC has assumed State communications responsibilities using Appendix F General Emergency Follow-up Information Form to communicate follow-up information.
[2]   Reevaluation of significant changes must additionally include a determination of whether Protective Action Recommendations (PARs) should be upgraded.
The need to upgrade PARs is determined through the continuous assessment of Appendix H Initial Protective Action Recommendations.
IF it has been determined that a PAR Upgrade is appropriate, THEN COMMUNICATE the Upgrade to the State using Appendix J Upgrade-Protective Action Recommendation.
[3]   CONTINUE to assess PARs until the CECC has assumed PAR responsibilities.
 
WBN                               General Emergency                                                       EPIP-5 Unit 0                                                                                                    Rev. 0043 Page 15 of 34 Appendix A (Page 1 of 1)
General Emergency Initial Notification Form
: 1. This is a Drill       This is an Actual Event - Repeat - This is an Actual Event
: 2. The SED at Watts Bar has declared a GENERAL EMERGENCY
: 3. EAL Designator:_______________________(Use only one EAL)
: 4. Radiological Conditions:       (Check one under both Airborne and Liquid column.)
Airborne Releases Offsite                           Liquid Releases Offsite Minor releases within federally approved limits1           Minor releases within federally approved Releases above federally approved limits1                 limits1 Release information not known                               Releases above federally approved limits1 (1Tech Specs/ODCM)                                           Release information not known (1Tech Specs/ODCM)
: 5. Event Declared:           Time:__________________                 Date:_____________________
: 6. The Meteorological Conditions are: (Use 46 meter data from the Met Tower. IF data is NOT available from the MET tower, contact the National Weather Service by dialing 9-1-423-586-8400. The National Weather Service will provide wind direction and wind speed.)
Wind Direction is FROM:_____________degrees                   Wind Speed: ______________m.p.h (15 minute average)                               (15 minute average)
: 7. Provide Protective Action Recommendation utilizing Appendix H: (Check either 1 or 2 or 3)
Recommendation 1                                       WIND            Recommendation 2 RECOMENDATION 1                 RECOMENDATION 2 FROM DEGREES EVACUATE LISTED SECTORS                              (Mark wind       EVACUATE LISTED SECTORS (2 mile Radius and 10 miles downwind)               direction          (2 mile Radius and 5 miles from step 6)          downwind)
SHELTER remainder of 10 mile EPZ                                       SHELTER remainder of 10 mile EPZ CONSIDER issuance of POTASSIUM                                                                                 CONSIDER issuance of IODIDE in accordance with the State Plan                                                                         POTASSIUM IODIDE in accordance with the State Plan A1, B1, C1, D1, C7, C9, D2, D4, D5, D6, D7, D8, D9                               From 26-68                       A1, B1, C1, D1, C7, D2, D4, D5 A1, B1, C1, D1, A3, A4, D2, D3, D4, D5, D6, D7, D8, D9                           From 69-110                     A1, B1, C1, D1, A3, D2, D4, D5 A1, B1, C1, D1, A2, A3, A4, A5, A6, A7, D2, D3, D5, D6                           From 111-170                     A1, B1, C1, D1, A2, A3, D2, D5 A1, B1, C1, D1, A2, A3, A5, A6, A7, B2, B3, B4, B5, C2                           From 171-230                     A1, B1, C1, D1, A2, A3, B2, B4, C2 A1, B1, C1, D1, B2, B3, B4, B5, C2, C3,                                         From 231-270                     A1, B1, C1, D1, B2, B4, C2 A1, B1, C1, D1, B2, B3, C2, C3, C4, C5, C6, C11                                 From 271-325                     A1, B1, C1, D1, B2, C2, C4, C5, A1, B1, C1, D1, C2, C4, C5, C6, C7, C8, C9, C10, C11, D4, D9                     From 326-25                     A1, B1, C1, D1, C2, C4, C5, C7, C8, D4 Recommendation 3 SHELTER all sectors CONSIDER issuance of POTASSIUM IODIDE in accordance with the State Plan Completed by (SED)______________ Peer Checked by ________________
 
WBN             General Emergency           EPIP-5 Unit 0                                      Rev. 0043 Page 30 of 34 Appendix H (Page 1 of 2)
Initial - Protective Action Recommendations
 
WBN             General Emergency           EPIP-5 Unit 0                                      Rev. 0043 Page 31 of 34 Appendix H (Page 2 of 2)
Initial - Protective Action Recommendations
 
WBN                       Site Area Emergency                   EPIP-4 Unit 0                                                          Rev. 0038 Page 15 of 28 Appendix A (Page 1 of 1)
Site Area Emergency Initial Notification Form
: 1.         This is a Drill         This is an Actual Event - Repeat - This is an Actual Event
: 2. The SED at Watts Bar has declared a Site Area Emergency
: 3. EAL Designator:           ________________________________
: 4. Radiological Conditions:         (Check one under both Airborne and Liquid column.)
Airborne Releases Offsite                             Liquid Releases Offsite Minor releases within federally approved             Minor releases within federally approved limits1                                               limits1 Releases above federally approved limits1             Releases above federally approved limits1 Release information not known                         Release information not known (1Tech Specs/ODCM)                                   (1Tech Specs/ODCM)
: 5. Event Declared:             Time:________________             Date:__________________
Eastern Time
: 6. Provide Protective Action Recommendation: None Completed By (SED) _______________________________________
Peer Checked By:           _______________________________________
 
WBN             Emergency Plan Classification Logic         EPIP-1 Unit 0                                                      Rev. 0042 Page 8 of 54 1.0     PURPOSE This Procedure provides guidance in determining the classification and declaration of an emergency based on plant conditions.
2.0     RESPONSIBILITY The responsibility of declaring an Emergency based on the guidance within this procedure belongs to the Shift Manager/Site Emergency Director (SM/SED) or designated Unit Supervisor (US) when acting as the SM or the TSC Site Emergency Director (SED). The following duties CAN NOT be delegated:
Emergency Classification, Emergency Dose Approval and PAR development prior to CECC Director ownership for PAR development.
3.0     INSTRUCTIONS 3.1     Precautions and Limitations CAUTION Unit-2 radiation monitor readings for classification purposes do not apply until Unit-2 is licensed and operating.
A. The criteria in WBN EPIP-1 are given for GUIDANCE ONLY: knowledge of actual plant conditions or the extent of the emergency may require that additional steps be taken. In all cases, this logic procedure should be combined with the sound judgment of the SM/SED and/or the TSC SED to arrive at a classification for a particular set of circumstances.
B. The Nuclear Power (NP) Radiological Emergency Plan (REP) will be activated when any one of the conditions listed in this logic is detected.
C. The SM/SED shall assess, classify, and declare an emergency condition within 15 minutes after information is first available to plant operators to recognize that an EAL has been exceeded and to make the declaration promptly upon identification of the appropriate Emergency Classification Level (ECL).
: 1. For EAL thresholds that specify duration of the off-normal condition, the emergency declaration process runs concurrently with the specified threshold duration.
: a. Consider as an example, the EAL fire which is not extinguished within 15 minutes of detection. On receipt of a fire alarm, the plant fire brigade is dispatched to the scene to begin fire suppression efforts.
 
CENTRAL EMERGENCY CONTROL CENTER (CECC)                                                             Page 6 of 44 OPERATIONS                           CECC EPIP-1                             Revision 60 3.3   CECC Director/Assistant CECC Director The CECC Director is responsible for directing TVA's overall response to the emergency.
An Assistant CECC Director (who is qualified as a CECC Director) may be used to assist the CECC Director in the accomplishment of position duties. The CECC Director, at his discretion, may delegate the accomplishment of duties to the Assistant CECC Director including signature authority.
The CECC Director ensures that Federal, State, and local agencies are notified in accordance with established procedures and that they are kept fully informed of all aspects of the emergency. The Director reviews with the Plant Assessment and Radiological Assessment Managers the onsite and offsite consequences of the accident and assesses the adequacy and need for measures taken for protection of the public. The Director coordinates TVA's efforts with State and Federal agencies involved in the offsite aspects of the emergency and requests any required federal assistance through the NRC.
Checklists for the CECC Director are provided in Appendices B through G. The CECC Director shall complete Appendix B for initial activation and the appropriate Appendix for the event (e.g.,
Appendix E for an Alert). After the appropriate level of CECC activation the CECC Director is responsible for the following:
* Approves all press releases developed in the CECC.
* Approves all press releases developed in the CECC.
* Notifies the appropriate state warning point of any emergency classification upgrades.
* Notifies the appropriate state warning point of any emergency classification upgrades.
* Notifies the appropriate state GAR of any emergency classification upgrades.
* Notifies the appropriate state GAR of any emergency classification upgrades.
* Approves and communicates any required Protective Action Recommendations (PARs) and PAR upgrades to the appropriate state warning point and GAR using Appendix I.
* Approves and communicates any required Protective Action Recommendations (PARs) and PAR upgrades to the appropriate state warning point and GAR using Appendix I.
* Maintains control of Safeguards Information within the CECC. 3.4 Plant Assessment Manager Plant Assessment Manager responsibilities are contained in CECC EPIP-6. 3.5 Radiological Assessment Manager Radiological Assessment Manager responsibilities are contained in CECC EPIP-7.
* Maintains control of Safeguards Information within the CECC.
3.6 Public Information Manager Public Information Manager responsibilities are contained in CECC EPIP-14.
3.4   Plant Assessment Manager Plant Assessment Manager responsibilities are contained in CECC EPIP-6.
WBN Unit 0 Emergency Exposure Guidelines EPIP-15 Rev. 0016 Page 14 of 16   Appendix C (Page 1 of 2) Emergency Respirator Issue Guidelines NOTE THESE GUIDELINES ARE RECOMMENDATIONS ONLY, SUBJECT TO THE JUDGEMENT OF RP AND EMERGENCY MANAGEMENT PERSONNEL. THESE GUIDELINES ARE APPLICABLE ONLY TO PROTECTION FROM AIRBORNE RADIOACTIVE MATERIAL AND DO NOT APPLY TO RESPIRATORS/SCBA'S ISSUED FOR PROTECTION FROM INDUSTRIAL OR CHEMICAL HAZARDS OR ATMOSPHERES IMMEDIATELY HAZARDOUS TO LIFE OR HEALTH. TASKS TO SAVE A LIFE OR PREVENT SIGNIFICANT  DAMAGE TO PLANT Respirator/SCBA not required to enter airborne radioactivity areas provided resulting internal dose plus external dose will not result in TEDE exceeding NRC dose limits or, if approved by the SED, doses up to the TVA emergency dose limits (i.e., up to 25 Rem/10 Rem) (this can include uptakes > 1 ALI)   HIGH PRIORITY TASKS (priority 1 or 2)
3.5   Radiological Assessment Manager Radiological Assessment Manager responsibilities are contained in CECC EPIP-7.
* Respirator/SCBA not required to enter airborne areas if the following are true: NOTE: IF INDIVIDUAL'S TOTAL INTAKE FOR THE YEAR TO DATE EXCEEDS 200 DAC-HRS., DOSE RESULTING FROM ALL INTAKES FOR THE YEAR TO DATE MUST BE ASSESSED IN DETERMINING THE TEDE.
3.6   Public Information Manager Public Information Manager responsibilities are contained in CECC EPIP-14.
* Individual's internal dose plus external dose will not result in TEDE exceeding NRC annual dose limit; and
 
* Delays or hindrances caused by issuing or wearing respirators/SCBAs will jeopardize the success or timeliness of the task; or
WBN               Emergency Exposure Guidelines           EPIP-15 Unit 0                                                      Rev. 0016 Page 14 of 16 Appendix C (Page 1 of 2)
* Use of a respirator/SCBA will result in a higher TEDE to the responding individuals. LOW or MID PRIORITY TASKS Use RCI-120 for respirator issue guidance. NOTE Protective requirements may be revised at the discretion of the TSC RP Manager as sample data becomes available.
Emergency Respirator Issue Guidelines NOTE THESE GUIDELINES ARE RECOMMENDATIONS ONLY, SUBJECT TO THE JUDGEMENT OF RP AND EMERGENCY MANAGEMENT PERSONNEL. THESE GUIDELINES ARE APPLICABLE ONLY TO PROTECTION FROM AIRBORNE RADIOACTIVE MATERIAL AND DO NOT APPLY TO RESPIRATORS/SCBAS ISSUED FOR PROTECTION FROM INDUSTRIAL OR CHEMICAL HAZARDS OR ATMOSPHERES IMMEDIATELY HAZARDOUS TO LIFE OR HEALTH.
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 6 of 16  C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)] Some examples of SRO exam items for this topic include:
TASKS TO SAVE A LIFE OR             Respirator/SCBA not required to enter airborne radioactivity PREVENT SIGNIFICANT                areas provided resulting internal dose plus external dose will DAMAGE TO PLANT                    not result in TEDE exceeding NRC dose limits or, if approved by the SED, doses up to the TVA emergency dose limits (i.e., up to 25 Rem/10 Rem) (this can include uptakes > 1 ALI)
HIGH PRIORITY TASKS
* Respirator/SCBA not required to enter airborne areas if the (priority 1 or 2)                  following are true:
NOTE:       IF INDIVIDUAL'S TOTAL
* Individual's internal dose plus external dose will not result INTAKE FOR THE YEAR    in TEDE exceeding NRC annual dose limit; and TO DATE EXCEEDS 200 DAC-HRS., DOSE
* Delays or hindrances caused by issuing or wearing RESULTING FROM ALL      respirators/SCBAs will jeopardize the success or timeliness INTAKES FOR THE YEAR    of the task; or TO DATE MUST BE ASSESSED IN
* Use of a respirator/SCBA will result in a higher TEDE to the DETERMINING THE TEDE. responding individuals.
LOW or MID PRIORITY TASKS           Use RCI-120 for respirator issue guidance.
NOTE Protective requirements may be revised at the discretion of the TSC RP Manager as sample data becomes available.
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]
Some examples of SRO exam items for this topic include:
* 10 CFR 50.59 screening and evaluation processes.
* 10 CFR 50.59 screening and evaluation processes.
* Administrative processes for temporary modifications.
* Administrative processes for temporary modifications.
* Administrative processes for disabling annunciators.
* Administrative processes for disabling annunciators.
* Administrative processes for the installation of temporary instrumentation.
* Administrative processes for the installation of temporary instrumentation.
* Processes for changing the plant or plant procedures. Section IV provides an example of a satisfactory SRO-only question related to this topic. D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.       [10 CFR 55.43(b)(4)] Some examples of SRO exam items for this topic include:
* Processes for changing the plant or plant procedures.
Section IV provides an example of a satisfactory SRO-only question related to this topic.
D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.
[10 CFR 55.43(b)(4)]
Some examples of SRO exam items for this topic include:
* Process for gaseous/liquid release approvals, i.e., release permits.
* Process for gaseous/liquid release approvals, i.e., release permits.
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures. Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits. SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc. E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
* Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedure's overall mitigative strategy or purpose.  
Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.
SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.
E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]
This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.
The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:
Page 6 of 16


The applicant's knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedure's content is required to correctly answer the written test item, for example: 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
 
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 7 of 16
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
* Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
* Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
SRO-only knowledge should not be claimed for questions that can be answered solely using "systems knowledge"; e.g.:
SRO-only knowledge should not be claimed for questions that can be answered solely using systems knowledge; e.g.:
* how the system works.
* how the system works.
* system flow path.
* system flow path.
* component locations, etc. SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:
* component locations, etc.
SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:
* the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
* the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
* any AOP entry condition.
* any AOP entry condition.
* plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
* plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
* immediate operator actions of a procedure. Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.        
* immediate operator actions of a procedure.
Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.
Page 7 of 16


Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) Page 8 of 16  Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5) (Assessment and selection of procedures)      
Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)
 
Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)
Can the question be answered solely by knowing "systems knowledge", i.e., how the system works, flowpath, logic, component location? RO question YesNoCan the question be answered solely by knowing immediate operator actions? YesCan the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?YesNoDoes the question require one or more of the following?
(Assessment and selection of procedures)
* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
Can the question be answered solely by knowing systems knowledge, i.e., how the system works,       Yes RO question flowpath, logic, component location?
* Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps
No Can the question be answered solely by knowing immediate operator actions?                           Yes    RO question No Can the question be answered solely by knowing entry conditions for AOPs or plant parameters         Yes RO question that require direct entry to major EOPs?
* Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
No Can the question be answered solely by knowing the purpose, overall sequence of events, or            Yes RO question overall mitigative strategy of a procedure?
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures SRO-only question Yes Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?NoNoYesQuestion might not be linked to 10 CFR 55.43(b)(5) for SRO-only NoRO question RO question RO question}}
No                                                          The SRO with the command function Does the question require one or more of the following?                 (US or Shift Manager) will
* Assessing plant conditions (normal, abnormal, or                   implement the emergency) and then selecting a procedure or section of a           REP. Note that the procedure to mitigate, recover, or with which to proceed           RO cannot perform
* Knowledge of when to implement attachments and                     this function.
appendices, including how to coordinate these items with procedure steps                                           Yes        SRO-only
* Knowledge of diagnostic steps and decision points in the               question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16}}

Latest revision as of 10:03, 5 February 2020

301 Draft SRO Written Exam
ML15259A292
Person / Time
Site: Watts Bar  Tennessee Valley Authority icon.png
Issue date: 09/16/2015
From:
Division of Reactor Safety II
To:
Tennessee Valley Authority
References
Download: ML15259A292 (321)
v • d • e


Text

NRC Exam Legend Times are HH:MM:SS [ 00:00:00 - 23:59:59 ]

ALARMS WINDOWS:

LIT DARK LIGHT INDICATIONS:

COLOR LIT DARK RED GREEN WHITE

76.

Given the following conditions:

- Unit 1 is at 28% power.

- RCP SEAL LEAK OFF FLOW HI (100-D) is LIT.

- 1-FR-62-24, SEAL LEAKOFF - HI RANGE - GPM reads 5.2 gpm.

Subsequently, the following is observed on 1-M-5:

In accordance with BOTH 1-AOI-24 and TI-12.04, which ONE of the following completes the statement below?

The Unit Supervisor will direct ____(1)____.

NOTE: 1-AOI-24, RCP Malfunctions During Pump Operation TI-12.04, Users Guide for Abnormal and Emergency Operating Instructions ES-0.1, Reactor Trip Response A. a Unit 1 shutdown in accordance with 1-GO-5, Unit Shutdown From 30%

Reactor Power to Hot Standby B. the OAC to trip the Unit 1 reactor and the CRO to perform further actions of 1-AOI-24 after the crew transitions to ES-0.1 C. the OAC to trip the Unit 1 reactor and the CRO to perform further actions of 1-AOI-24 while the OAC is performing his immediate actions D. the crew to maintain Unit 1 at power as a Unit Shutdown is NOT yet required and an operator to REFER TO Attachment 2, Immediate Shutdown Criteria

CORRECT ANSWER: B DISTRACTOR ANALYSIS:

A. Incorrect: This distractor is incorrect and is tantamount in plausibility as the D distractor. If the US entered section 3.3 of 1-AOI-24 and then believed that seal leakoff flow were in excess of the criteria given for an initial controlled shutdown, then he would select an appropriate procedure (again because the unit was at a power of less than 30%,

1-GO-5 would be appropriate) and bring the Unit to mode 3. Therefore, the foregoing defends the plausibility of this distractor as this would be a correct procedural flowpath if the applicant did not understand that a trip setpoint were exceeded and that the seal leakoff criteria were exceeded.

B. Correct: As mentioned, this distractor lists the correct action for the control room staff to take; this is, tripping the reactor when an immediate shutdown criterion is met and continuing with the steps of 1-AOI-24 after the transition to ES-0.1 is made.

C. Incorrect: It is correct that a reactor trip would be required as an immediate shutdown criterion had been exceeded. Even if the Unit Supervisor did not initially recognize that such criteria had been exceeded and thus went to section 3.3; he would reach a step in such section which details MONITOR RCP immediate shutdown required. At this point he would enter section 3.2 and direct a reactor trip. Section 2.8, Use of AOIs While in EOIs of TI-12.04 contains: 3. When an AOI in effect directs a Reactor Trip and then the performance of the AOI should continue immediately following transition to ES-0.1. Therefore, it would not be correct for the CRO to perform the steps of 1-AOI-24 in parallel with the immediate actions of the OAC (because the transition to ES-0.1 had not yet been made). It is plausible to believe this because if one did not recognize the restriction imposed by TI-12.04, one would logically interpret step 3. of 1-AOI-24 as directing exactly this. Step 3 of 1-AOI-24 reads: TRIP the reactor, and GO TO E-0 Reactor Trip or Safety Injection, WHILE continuing with this instruction.

D. Incorrect: As seen in ARI-95-101, Reactor Coolant Pumps, the setpoint for annunciator window 100-D is 4.8 gpm. Plant Operation has shown that leakoff values of approximately 2.3 gpm (this value is slightly variable) are normal. The question gives the fact that leakoff for the #1 seal for the #1 RCP is 5.2 gpm. Using Attachment 1 of 1-AOI-24, one may observe that the normal operating range for the

  1. 1 seal leakoff is between 1 to 5 gpm when the Unit is at normal operating pressure of 2235 psig. Therefore, it is fact that #1 seal leakoff is high. The stem of the question also gives the fact that 1-TI-62-3; RCP 1 LWR BRG TEMP is at 230°F.

This is in excess of immediate trip criteria of 225°F (contained in Attachment 2 of 1-AOI-24). Two correct procedural avenues could be used to address this issue. Both will result in the same outcome. Firstly, the Unit Supervisor could immediately identify that RCP immediate trip criteria are met through memorization of the criteria of Attachment 2, RCP Immediate Shutdown Criteria and thus upon entry into 1-AOI-24 would select subsection 3.2, RCP Tripped or Shutdown Required. If the Unit Supervisor did not immediately identify that a trip of the RCP were required he would select section 3.3, #1 Seal Leakoff Flow High. Section 3.3 contains a decision step which selects whether a controlled shutdown (given that the unit is at 28% power a shutdown conducted per 1-GO-5 would be appropriate) is initially appropriate. If #1 seal leakoff is greater than or equal to 6.0 gpm then an initial shutdown is performed. If the seal leakoff is not in excess of this value, then the Unit

Supervisor would assign an operator to REFER TO Attachment 2 and thus utilize the Attachment to monitor for further degradation of the RCP seal package. If the Unit Supervisor did not recognize that an immediate trip of the RCP was required and because he had already bypassed the opportunity for plant shutdown (because seal leakoff were less than 6.0 gpm) then he would continue performing section 3.3 and thus maintain the Unit at power. The foregoing defends the plausibility of this distractor as this would be a correct procedural flowpath if the applicant did not understand that a trip setpoint were exceeded.

Question Number: 76 Tier: 1 Group: 1 K/A: 015/017 Reactor Coolant Pump (RCP) Malfunctions 2.2 Equipment Control 2.2.44 Ability to interpret control room indications to verify the status and operation ofl a system, and understand how operator actions and directives affect plant and system conditions.

Importance Rating: 4.2 4.4 10 CFR Part 55: (CFR: 41.5 / 43.5 / 45.12) 10CFR55.43.b: 10 CFR 55.43(b)(5)

K/A Match: K/A is matched because the applicant is required to interpret the #1 RCP parameters to verify that the RCP is operating improperly and thus that the correct operator directives will cause the plant to be tripped and the RCP to be secured in accordance with the guidance of 1-AOI-24.

Technical

Reference:

ARI-95-101, Reactor Coolant Pumps 1-AOI-24, RCP Malfunctions During Pump Operation 0-TI-12.04, Users Guide for Abnormal and Emergency Operating Instructions Proposed references to None be provided:

Learning Objective: 3-OT-AOI2400

5. Given a set of plant conditions, DESCRIBE operator actions required in response to the following per AOI-24, RCP Malfunctions during Pump Operation:
a. RCP tripped or shutdown required
b. #1 Seal Leakoff Flow HIGH
c. #1 Seal Leakoff Flow LOW AND Standpipe level alarm DARK,
d. #2 Seal Leakoff Flow HIGH
e. #3 Seal Leakoff Flow HIGH Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question for the 2015-301 NRC SRO Exam Comments: The question is SRO only as this question meets the general SRO only criteria of Assessment of facility conditions an selection of appropriate procedures during normal, abnormal, and emergency situations.

WBN RCP MALFUNCTIONS DURING PUMP 1-AOI-24 Unit 1 OPERATION Rev. 0000 3.0 OPERATOR ACTION:

3.1 Diagnostics IF GO TO Subsection RCP tripped or shutdown required 3.2

  1. 1 seal leakoff flow HIGH, 3.3
  1. 1 seal leakoff flow LOW, 3.4 AND Seal leakoff is high (at 5.2 gpm). It is Standpipe level alarm DARK, plausible to enter this section if one
  1. 2 Seal Leakoff Flow HIGH 3.5 did not (upon entry into this procedure)

(#1 seal leakoff flow LOW, recognize that an immediate AND shutdown were Standpipe level alarm LIT), required.

  1. 3 seal leakoff flow HIGH 3.6

(#1 seal leakoff flow NORMAL AND Standpipe level alarm LIT),

Page 5 of 27

WBN RCP MALFUNCTIONS DURING PUMP 1-AOI-24 Unit 1 OPERATION Rev. 0000 Step Action/Expected Response Response Not Obtained 3.3 # 1 Seal Leakoff Flow High CAUTION A seal leakoff rise to greater than 2.0 gpm AFTER experiencing low leakoff of less than 0.8 gpm may indicate seal degradation.

Plant Management should be notified of leakoff trends.

NOTE 1 Anytime #1 seal leakoff flow exceeds the values shown on Attachment 1, system engineering should be requested to perform an evaluation of the #1 seal condition.

NOTE 2 During plant startup after seal maintenance, the #1 seal may require 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of run time before the seal seats fully and operates normally.

NOTE 3 The #1 seal return should be isolated between 3 and 5 minutes after tripping an RCP to allow for pump coastdown.

1. MONITOR #1 seal leakoff equal to or **GO TO Step 5.

greater than 6.0 gpm.

2. MONITOR RCPs lower bearing and **GO TO Subsection 3.2, Step 2.
  1. 1 seal outlet temp STABLE or DROPPING.
3. REFER TO appropriate instruction to initiate a controlled shutdown to Mode 3 while continuing with this instruction:
  • AOI-39, Rapid Load Reduction.
  • GO-4, Normal Power Operation.
  • GO-5, Unit Shutdown From 30%

Reactor Power to Hot Standby.

Page 11 of 27

WBN RCP MALFUNCTIONS DURING PUMP 1-AOI-24 Unit 1 OPERATION Rev. 0000 Step Action/Expected Response Response Not Obtained 3.3 # 1 Seal Leakoff Flow High (continued)

NOTE RCP shutdown time is based on an orderly reactor shutdown and may be delayed or expedited based on ongoing evaluations of current plant conditions, other pump parameters and efforts to restore seal leakoff flows to normal.

4. REMOVE RCP from service:
  • Within 8 hrs, OR
  • As directed by Plant Management.
5. MONITOR RCP immediate ** GO TO Step 6.

shutdown required:

  • REFER TO ATTACHMENT 2, RCP Immediate Shutdown Criteria.
    • GO TO Subsection 3.2, Step 2.
6. ADJUST seal injection flow to exceed total #1 seal leakoff rate.
7. CONTACT System Engineer for further guidance WHILE continuing this Instruction:
  • Recommendations for continued RCP operation.
  • Installation of alternate flow measuring equipment (flows greater than 6 gpm).

Page 12 of 27

WBN RCP MALFUNCTIONS DURING PUMP 1-AOI-24 Unit 1 OPERATION Rev. 0000 Step Action/Expected Response Response Not Obtained 3.3 # 1 Seal Leakoff Flow High (continued)

CAUTION If all RCP seal cooling is lost, cooling down and depressurizing the RCS at a rapid rate, within established guidelines will minimize seal leakage.

8. CHECK seal injection flow between ADJUST 1-HIC-62-89A and 1-HIC-62-93A 8 and 13 gpm/RCP. to establish seal injection flow between 8 and 13 gpm/RCP.

IF seal injection remains less than 8 gpm/RCP, THEN:

a. ENSURE CCS flow to thermal barrier.
b. ENSURE RCP pump lower bearing and #1 seal outlet remains less than 225°F.
c. EVALUATE changing seal injection filter(s).
9. CONTROL VCT outlet temp less than 123°F:
  • ADJUST 1-HIC-62-78A.
  • ADJUST charging and letdown flow to reduce regenerative heat-exchanger outlet temp.
10. CHECK VCT pressure between 15 ADJUST VCT pressure:

and 30 psig.

  • VENT VCT by controlling 1-FCV-62-125, OR
  • CONTROL VCT level by diversion or makeup.

Page 13 of 27

WBN RCP MALFUNCTIONS DURING PUMP 1-AOI-24 Unit 1 OPERATION Rev. 0000 Step Action/Expected Response Response Not Obtained 3.3 # 1 Seal Leakoff Flow High (continued)

11. MONITOR RCP lower bearing and IF temp greater than 180°F AND rising,
  1. 1 seal outlet temp: THEN
  • Less than or equal to 180°F ** GO TO Subsection 3.2.
  • STABLE or DROPPING.
12. INITIATE repairs as required.
13. RETURN TO Instruction in effect.

End of Section Page 14 of 27

WBN RCP MALFUNCTIONS DURING PUMP 1-AOI-24 Unit 1 OPERATION Rev. 0000 Attachment 2 (Page 1 of 1)

RCP IMMEDIATE SHUTDOWN CRITERIA NOTE Exceeding any of the following setpoints will require an immediate pump shutdown. Operating limits can be found in SOI 68.02. This list is immediate shutdown criteria only.

A. Shaft vibration greater than 20 mils or 15 mils with a rate of rise equal to 1 mil/hr (alarm at 15 mils). [Indicators located on 0-PNL-52-R139, Aux Inst Rm.]

B. Frame vibration greater than 5 mils or 3 mils with a rate of rise of 0.2 mil/hr.

[Readings taken by Maint. at Aux Bldg L-Panels, el.737.]

C. Motor windings temp greater than 302°F.

D. Motor bearing temp greater than 195°F.

E. Pump bearing temp greater than 225°F.

F. Loss of CCS to oil coolers for greater than 10 minutes.

G. No. 1 seal outlet temp greater than 225°F.

H. No. 1 seal flow HIGH with rising pump bearing or #1 seal leakoff temperatures.

I. No. 1 seal P less than or equal to 200 psid.

Page 26 of 27

WBN User's Guide for 0-TI-12.04 Unit 1 & 2 Abnormal and Emergency Rev. 0000 Operating Instructions Page 35 of 57 2.7 Prudent Operator Actions (continued)

3. The operator should consult nearby personnel who are suitably qualified and notify them of their proposed actions. If no disagreement is forthcoming, he should then take the necessary mitigation or preemptive actions to terminate the event.
4. The STAR principle should be applied --Stop, Think, Act, Review. Ask yourself: If I take this action, could I inadvertently cause other more severe problems? Am I better off taking no action at all? How will safety status be affected?

2.8 Use of AOIs While in EOIs

1. During performance of the 1-ES-0.1, if plant conditions warrant implementation of an AOI, then the required AOI may be performed concurrently (on a not-to-interfere basis) with the EOIs.
2. When running an AOI concurrently with an EOI (1-ECA-0.0, 1-ES-0.1, etc.)

the Unit Supervisor/SRO will assign the BOP/CRO operator responsibility for the AOI if another Unit Supervisor is NOT available. If the BOP/CRO operator performs an AOI, he/she should consult directly with the Unit Supervisor and give them the status as required by the AOI.

3. When an AOI in effect directs a Reactor Trip, then the performance of the AOI should continue immediately following transition to 1-ES-0.1.

Performance assignments will be at the discretion of the SM/US based on the status and importance of events in progress.

4. When implementing an AOI outside the horseshoe in the control room, the Unit Supervisor should accompany the board operator to read the procedure steps and direct actions of the operator, unless higher priority conditions demanding the Unit Supervisors attention exist; in which case the BOP/CRO should implement the AOI using the single performer method. The actively licensed STA may serve as a reader unless the crew is in progress of performing actions within the EOI network.

3.0 RECORDS None

WBN RCP MALFUNCTIONS DURING PUMP 1-AOI-24 Unit 1 OPERATION Rev. 0000 3.0 OPERATOR ACTION:

3.1 Diagnostics IF GO TO Subsection RCP tripped or shutdown required 3.2

  1. 1 seal leakoff flow HIGH, 3.3
  1. 1 seal leakoff flow LOW, 3.4 AND Standpipe level alarm DARK,
  1. 2 Seal Leakoff Flow HIGH 3.5

(#1 seal leakoff flow LOW, AND Standpipe level alarm LIT),

  1. 3 seal leakoff flow HIGH 3.6

(#1 seal leakoff flow NORMAL AND Standpipe level alarm LIT),

Page 5 of 27

WBN RCP MALFUNCTIONS DURING PUMP 1-AOI-24 Unit 1 OPERATION Rev. 0000 Step Action/Expected Response Response Not Obtained 3.2 RCP Tripped Or Shutdown Required NOTE 1 Malfunctions addressed by this procedure require RCP shutdown as soon as possible.

NOTE 2 Exceeding any of the limits listed on Attachment 2 of this procedure will require immediate shutdown of the affected RCP.

NOTE 3 Malfunctions resulting in high #1 seal leakoff will require closing #1 seal return FCV following RCP coastdown

1. CHECK RCP tripped MONITOR RCP immediate shutdown Criteria:
  • REFER TO ATTACHMENT 2, RCP Immediate Shutdown Criteria.
1) IF RCP immediate shutdown required, THEN
    • GO TO Step 2.
2) IF RCP immediate shutdown NOT required, THEN
    • GO TO Step 9
2. CHECK unit in Mode 1 or 2 ** GO TO Step 4.

Page 6 of 27

WBN RCP MALFUNCTIONS DURING PUMP 1-AOI-24 Unit 1 OPERATION Rev. 0000 Step Action/Expected Response Response Not Obtained 3.2 RCP Tripped Or Shutdown Required (continued)

NOTE Control room staff should brief on Steps 3 through 6 prior to tripping the reactor. This ensures that the affected RCP is stopped and that appropriate actions are taken when unit is removed from service.

3. TRIP the reactor, and .

GO TO E-0, Reactor Trip or Safety Injection, WHILE continuing with this instruction.

4. STOP and LOCK OUT affected RCP(s).
5. IF in Mode 3, **GO TO ES-0.2, Natural Circulation THEN Cooldown, WHILE continuing with this CHECK any RCP Running instruction CAUTION If the RCP seal return flow control valve (FCV) is NOT closed within 5 minutes of stopping the RCP with excessive leakoff, seal damage may occur.
6. MONITOR RCP seal leakoff less WHEN the RCP has coasted down than 6 gpm per pump: (between 3 and 5 minutes),

THEN

  • 1-FR-62-24 [RCP 1 & 2]
  • 1-FR-62-50 [RCP 3 & 4] CLOSE affected RCP seal return FCV:
  • 1-FCV-62-9 [RCP 1]
  • 1-FCV-62-22 [RCP 2]
  • 1-FCV-62-35 [RCP 3]
  • 1-FCV-62-48 [RCP 4]
7. CHECK RCPs 1 and 2 running. CLOSE affected loops pressurizer spray valve.

Page 7 of 27

WBN RCP MALFUNCTIONS DURING PUMP 1-AOI-24 Unit 1 OPERATION Rev. 0000 Step Action/Expected Response Response Not Obtained 3.2 RCP Tripped Or Shutdown Required (continued)

8. GO TO Step 15.
9. CONSULT plant staff as necessary for recommendations for continued RCP operation.

NOTE Control room staff should brief on Steps 10 through 13 prior to reducing load. This ensures that the affected RCP is stopped and that appropriate actions are taken when unit is removed from service.

10. IF removal of RCP(s) is required, RETURN TO instruction in effect.

THEN REFER TO appropriate instruction to initiate a controlled shutdown to Mode 3 while continuing with this instruction:

  • AOI-39, Rapid Load Reduction
  • GO-4, Normal Power Operation.
  • GO-5, Unit Shutdown From 30%

Reactor Power to Hot Standby

11. MAINTAIN affected SG level on PROGRAM:
  • LOWER MFW flow as steam flow drops.
  • ISOLATE blowdown from affected SG.
12. WHEN unit is in Mode 3, THEN Page 8 of 27

WBN RCP MALFUNCTIONS DURING PUMP 1-AOI-24 Unit 1 OPERATION Rev. 0000 Step Action/Expected Response Response Not Obtained 3.2 RCP Tripped Or Shutdown Required (continued)

a. STOP and LOCK OUT affected RCP(s).
b. CHECK any RCP Running b. **GO TO ES-0.2, Natural Circulation Cooldown, WHILE continuing with this instruction.

CAUTION If the RCP seal return flow control valve (FCV) is NOT closed within 5 minutes of stopping the RCP with excessive leakoff, seal damage may occur.

13. MONITOR RCP seal leakoff less than WHEN the RCP has coasted down 6 gpm per pump: (between 3 and 5 minutes),

THEN

  • 1-FR-62-24 [RCP 1 & 2]

CLOSE affected RCP seal return FCV:

  • 1-FR-62-50 [RCP 3 & 4]
  • 1-FCV-62-9 [RCP 1]
  • 1-FCV-62-22 [RCP 2]
  • 1-FCV-62-35 [RCP 3]
  • 1-FCV-62-48 [RCP 4]
14. CHECK RCPs 1 and 2 running. CLOSE affected loops pressurizer spray valve.

Page 9 of 27

WBN RCP MALFUNCTIONS DURING PUMP 1-AOI-24 Unit 1 OPERATION Rev. 0000 Step Action/Expected Response Response Not Obtained 3.2 RCP Tripped Or Shutdown Required (continued)

15. REFER TO Tech Spec:
16. INITIATE repairs as required.
17. OBTAIN plant management approval prior to restarting any RCP.
18. RETURN TO Instruction in effect.

End of Section Page 10 of 27

WBN Reactor Coolant Pumps ARI-95-101 Unit 1 Rev. 0035 Page 39 of 50 100-D Source Setpoint RCP 1: 1-FS-62-11 4.8 gpm RCP SEAL LEAK OFF RCP 2: 1-FS-62-24 FLOW RCP 3: 1-FS-62-37 HI RCP 4: 1-FS-62-50 (Page 1 of 1)

Probable A. No. 1 seal damage Cause: B. No. 1 seal NOT fully seated C. Loss of seal injection water followed by high seal temperature Corrective [1] VERIFY high leakoff flow condition of affected RCP(s) with the following Action: instruments:

RCP RECORDER PEN/TRACE ICS POINT 1 1-FR-62-24 Red F1018A 2 1-FR-62-24 Blue F1020A 3 1-FR-62-50 Red F1022A 4 1-FR-62-50 Blue F1024A

[2] IF high leakoff is confirmed, THEN GO TO AOI-24, RCP MALFUNCTIONS DURING PUMP OPERATION.

References:

1-47W610-62-1 AOI-24

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]

Some examples of SRO exam items for this topic include:

  • Administrative processes for the installation of temporary instrumentation.
  • Processes for changing the plant or plant procedures.

Section IV provides an example of a satisfactory SRO-only question related to this topic.

D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

[10 CFR 55.43(b)(4)]

Some examples of SRO exam items for this topic include:

  • Process for gaseous/liquid release approvals, i.e., release permits.
  • Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.

Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.

SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.

E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.

The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:

Page 6 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

  • Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
  • Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
  • Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.

SRO-only knowledge should not be claimed for questions that can be answered solely using systems knowledge; e.g.:

  • how the system works.
  • system flow path.
  • component locations, etc.

SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:

  • the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
  • any AOP entry condition.
  • plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
  • immediate operator actions of a procedure.

Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.

Page 7 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, Yes RO question flowpath, logic, component location?

No Can the question be answered solely by knowing immediate operator actions? Yes RO question No Can the question be answered solely by knowing entry conditions for AOPs or plant parameters Yes RO question that require direct entry to major EOPs?

No Can the question be answered solely by knowing the purpose, overall sequence of events, or Yes RO question overall mitigative strategy of a procedure?

No Does the question require one or more of the following?

  • Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
  • Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps Yes SRO-only
  • Knowledge of diagnostic steps and decision points in the question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
  • Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

77.

Given the following conditions:

- 1-FR-S.1, Nuclear Power Generation/ATWS was entered.

- The Unit Supervisor has reached step 12, MONITOR reactor subcriticality:

- The following indications are noted:

Which ONE of the following completes the statements listed below?

In accordance with 1-FR-S.1, the conditions shown above _____(1)_____ CURRENTLY allow emergency boration to be terminated.

In accordance with the Westinghouse background document for 1-FR-S.1, Step 4, INITIATE RCS Boration: _____(2)_____ a TIME CRITICAL step.

A. (1) does (2) is B. (1) does (2) is NOT C. (1) does NOT (2) is D. (1) does NOT (2) is NOT

CORRECT ANSWER: D DISTRACTOR ANALYSIS:

A. Incorrect: Step 12 of 1-FR-S.1 states: MONITOR reactor subcriticality: a. CHECK Power range channels less than 5%. b. CHECK Intermediate range startup rate NEGATIVE. The conditions displayed in the stem of the question indicate that the PRNIs are at 4% but that the IR SUR is 0 dpm (i.e. not negative). Therefore, the procedure user is directed to continue in 1-FR-S.1 and not terminate emergency boration. It is plausible to believe that an IR SUR of 0 would allow the termination of emergency boration because the status tree for subcriticality allows for a ZERO SUR as a check for reactor subcriticality (in the decision tree INTERMEDIATE RANGE SUR ZERO OR NEGATIVE).

While the Westinghouse background document for 1-FR-S.1 states that Emergency Boration of the RCS is the most direct manner of adding negative reactivity to the core, it does not regard this step as time critical. The foregoing supports the plausibility for the belief that Emergency Boration is time critical.

B. Incorrect: Again, it is incorrect and yet plausible that the conditions shown in the stem of the question do allow emergency boration to be terminated. Also, it is correct that the initiation of Emergency Boration is not a time critical step.

C. Incorrect: It is correct that the conditions displayed do NOT allow the boration to be terminated. It is incorrect and yet plausible that the initiation of boration is a time critical step.

D. Correct: It is correct that the conditions displayed do NOT allow the boration to be terminated. It is correct that the initiation of boration is NOT a time critical step.

Question Number: 77 Tier: 1 Group: 1 K/A: 029 Anticipated Transient Without Scram (ATWS)

EA2 Ability to determine or interpret the following as they apply to a ATWS:

EA2.01 Reactor nuclear instrumentation Importance Rating: 4.4 4.7 10 CFR Part 55: (CFR 43.5 / 45.13) 10CFR55.43.b: 10 CFR 55.43(b)(5)

K/A Match: K/A is matched because the applicant is required to interpret both the power range NI readings as well as the intermediate range startup rate indications as they apply to an ATWS. The applicant must then decide if the sub criticality criteria are met and the subsequent required procedural actions.

Westinghouse Background Document for 1-FR-S.1 Technical

Reference:

1-FR-S.1, Nuclear Power Generation/ATWS FR-0, Status Trees Proposed references to None be provided:

Learning Objective: 3-OT-FRS0001

9. Given a set of plant conditions, use 1-FR-S.1, FR-S.2 and the Critical Safety Function Status Trees to correctly DIAGNOSE and implement:

Action Steps, RNOs, Notes and Cautions.

10. EXPLAIN the purpose for and basis of each step in 1-FR-S.1 and FR-S.2 Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question for the 2015-301 NRC SRO Exam Comments: The question is SRO only as this question meets the general SRO only criteria of Assessment of facility conditions an selection of appropriate procedures during normal, abnormal, and emergency situations.

WBN Status Trees FR-0 Unit 1 Rev. 0014 Attachment 1 (Page 1 of 8)

Monitoring Critical Safety Functions SUBCRITICALITY FR-S Page 4 of 11

WBN Nuclear Power Generation/ATWS 1-FR-S.1 Unit 1 Rev. 0001 Step Action/Expected Response Response Not Obtained

9. ENSURE the following trips:
a. Reactor Trip. a. DISPATCH operator to locally trip reactor:

[MG set room].

  • OPEN breakers to MG sets

[480V unit boards A and B].

b. Turbine Trip. b. DISPATCH operator to locally trip turbine:
  • TRIP from front standard.
  • STOP and PULL TO LOCK both EHC pumps.
10. MAINTAIN rod insertion UNTIL rods fully inserted.
11. REFER TO EPIP-1, Emergency Plan Classification Flowchart for ATWS event.
12. MONITOR reactor subcriticality:
a. CHECK Power range channels a. ** GO TO Step 13.

less than 5%.

b. CHECK Intermediate range b. ** GO TO Step 13.

startup rate NEGATIVE.

c. ** GO TO Step 21. POWER MUST BE LESS THAN 5%

and SUR must be NEGATIVE.

step 21 is terminate This is different emergency boration than the status trees which list IR Page 6 of 16 SUR is 0.

WBN Nuclear Power Generation/ATWS 1-FR-S.1 Unit 1 Rev. 0001 Step Action/Expected Response Response Not Obtained

19. CHECK Incore T/Cs less than IF Incore T/Cs are greater than 1200°F 1200°F. AND rising, THEN
    • GO TO 1-SACRG-1, Severe Accident Control Room Guideline Initial Response.
20. CHECK reactor subcritical: CONTINUE to borate.
a. Power range channels less than IF boration is NOT available, 5%. THEN ALLOW RCS to heat up to insert
b. Intermediate range startup rate negative reactivity from temperature NEGATIVE.

coefficients.

IF red OR orange condition exists on other Status Trees, THEN PERFORM actions of other FR Procedures which do not cool down or otherwise add positive reactivity to the core.

    • GO TO Step 4.
21. TERMINATE emergency boration:
a. PLACE BA transfer pumps in SLOW speed.
b. CLOSE emergency borate valve 1-FCV-62-138.
c. IF alternate boration opened, THEN Locally CLOSE 1-ISV-62-929.

Page 10 of 16

STEP DESCRIPTION TABLE FOR FR-S.l Step __2__

STEP: Verify Turbine Trip This is a time critical step (within PURPOSE: To ensure that the turbine is tripped 30 seconds).

BASIS:

The turbine is tripped to prevent an uncontrolled cool down of the RCS due to steam flow that the turbine would require. For an ATWS event where a loss of normal feedwater has occurred, analyses have shown that a turbine trip is necessary (within 30 seconds) to maintain SG inventory.

If the turbine will not trip, a turbine runback (manual decrease in load) at maximum rate will also reduce steam flow in a delayed manner. If the turbine stop valves cannot be closed by either trip or runback, the MSIVs should be closed.

ACTIONS:

o Determine if all turbine stop valves are closed o Determine if turbine will not trip o Determine if turbine cannot be run back o Trip the turbine o Manually run back turbine o Close main steamline isolation and bypass valves INSTRUMENTATION:

o Turbine stop valve position indication o MSIVs and bypass valves position indication CONTROL/EQUIPMENT:

o Switches for turbine trip (e.g. manual trip buttons, overspeed test switch, EH control oil pump switches) o Controls to manually run back turbine o Switches to close MSIVs and bypass valves FR-S.l Background 77 HP-Rev. 2, 4/30/2005 HFRSIBG.doc

STEP DESCRIPTION TABLE FOR FR-S.l Step 3__

STEP: Check AFW Pumps Running This is another time critical step PURPOSE: To ensure AFW pumps are running (within 60 seconds).

BASIS:

The MD AFW pumps start automatically on an SI signal and SG low level to provide feed to the SGs for decay heat removal. If SG levels drop below the appropriate setpoint, the turbine-driven AFW pump will also automatically start to supplement the MD pumps. The ATWS analyses have shown that actuation of AFW within 60 seconds after the failure to scram provides acceptable results.

ACTIONS:

o Determine if MD AFW pumps are running o Determine if the turbine-driven AFW pump is running if necessary o Start MD AFW pumps o Open steam supply valves to turbine-driven AFW pump INSTRUMENTATION:

o MD AFW pumps status indication o Turbine-driven AFW pump status indication o Turbine-driven AFW pump steam supply valve position indication CONTROL/EQUIPMENT:

Switches for:

o MD AFW pumps o Turbine-driven AFW pump steam supply valves KNOWLEDGE:

N/A PLANT-SPECIFIC INFORMATION:

N/A FR-S.l Background 79 HP-Rev. 2, 4/30/2005 HFRSIBG.doc

Notice that even though this is "the most direct manner of adding negative STEP DESCRIPTION TABLE FOR FR-S.l reactivity Step 4__ to the core," it is not a time critical step.

STEP: Initiate Emergency Boration of RCS PURPOSE: To add negative reactivity to bring the reactor core subcritical BASIS:

After control rod trip and rod insertion functions, boration is the next most direct manner of adding negative reactivity to the core. The intended boration path here is the most direct one available, not requiring SI initiation, but using normal charging pump(s). Pump miniflow lines are assumed to be open to protect the pumps in the event of high RCS pressure.

Several plant specific means are usually available for rapid boration and should be specified here in order of preference. Methods of rapid boration include emergency boration, injecting the BIT, and safety injection actuation.

It should be noted that SI actuation will trip the main feedwater pumps. If this is undesirable, the operator can manually align the system for safety injection. However, the RWST valves to the suction of the SI pumps should be opened first before opening up the BIT valves. If a safety injection is already in progress but is having no effect on nuclear flux, then the BIT and RWST are not performing their intended function, perhaps due to blockage or leakage. In this case some other alignment using the BATs and/or non-safeguards charging pump(s) is required.

The check on RCS pressure is intended to alert the operator to a condition which would reduce charging or SI pump injection into the RCS and, therefore, boration. The PRZR PORV pressure setpoint is chosen as that pressure at which flow into the RCS is insufficient. The contingent action is a rapid depressurization to a pressure which would allow increased injection flow.

When primary pressure drops 200 psi below the PORV pressure setpoint, the PORVs should be closed. The operator must verify successful closure of the PORVs, closing the isolation valves, if necessary.

FR-S.l Background 80 HP-Rev. 2, 4/30/2005 HFRSIBG.doc

STEP DESCRIPTION TABLE FOR FR-S.l Step 4__

ACTIONS:

o Determine if PRZR pressure is less than (A.02) psig o Determine if PRZR PORVs and block valves are open o Start charging/SI pumps o Start PO pump o Align boration path o Align charging flow path o Open PRZR PORVs and block valves as necessary until PRZR pressure is less than (A.08) psig INSTRUMENTATION:

o Charging/SI pump(s) status indication o PO pump status indication o Position indication for charging path valves, boration path valves o PRZR pressure indication o PRZR PORV and block valve position indications CONTROL/EQUIPMENT:

o Charging/SI pump(s) switches o PO pump switch o Switches for charging path valves/boration path valves o PRZR PORVs and block valves switches KNOWLEDGE:

N/A PLANT-SPECIFIC INFORMATION:

o (A.02) PRZR PORV pressure setpoint.

o (A.08) 200 psi less than PRZR PORV pressure setpoint.

o Preferred alignments for emergency boration based on plant equipment and operating practices.

FR-S.1 Background 81 HP-Rev. 2, 4/30/2005 HFRS1BG.doc

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]

Some examples of SRO exam items for this topic include:

  • Administrative processes for the installation of temporary instrumentation.
  • Processes for changing the plant or plant procedures.

Section IV provides an example of a satisfactory SRO-only question related to this topic.

D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

[10 CFR 55.43(b)(4)]

Some examples of SRO exam items for this topic include:

  • Process for gaseous/liquid release approvals, i.e., release permits.
  • Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.

Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.

SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.

E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.

The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:

Page 6 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

  • Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
  • Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
  • Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.

SRO-only knowledge should not be claimed for questions that can be answered solely using systems knowledge; e.g.:

  • how the system works.
  • system flow path.
  • component locations, etc.

SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:

  • the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
  • any AOP entry condition.
  • plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
  • immediate operator actions of a procedure.

Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.

Page 7 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, Yes RO question flowpath, logic, component location?

No Can the question be answered solely by knowing immediate operator actions? Yes RO question No Can the question be answered solely by knowing entry conditions for AOPs or plant parameters Yes RO question that require direct entry to major EOPs?

No Can the question be answered solely by knowing the purpose, overall sequence of events, or Yes RO question overall mitigative strategy of a procedure?

No Does the question require one or more of the following?

  • Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
  • Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps Yes SRO-only
  • Knowledge of diagnostic steps and decision points in the question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
  • Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

78.

Given the following conditions:

- The Unit 1 #1 SG is ruptured.

- The RCPs are SECURED.

- In accordance with step 18 of 1-E-3, the crew has INITIATED RCS cooldown to the target incore temperature.

- An ORANGE path exists for 1-FR-P.1 based SOLELY upon the Tcold in Loop #1.

Which ONE of the following describes the procedure transition requirements in accordance with 1-E-3?

The Unit Supervisor WILL __________.

NOTE: 1-E-3, Steam Generator Tube Rupture 1-FR-P.1, Pressurized Thermal Shock A. IMMEDIATELY transition to 1-FR-P.1 B. NOT transition to 1-FR-P.1 UNTIL 1-E-3 is completed C. transition to 1-FR-P.1 IF the ORANGE path still exists ONCE the cooldown to target incore temperature is completed D. transition to 1-FR-P.1 IF the ORANGE path still exists ONCE SI is terminated in accordance with 1-E-3

CORRECT ANSWER: D DISTRACTOR ANALYSIS:

A. Incorrect: As seen in 1-E-3, 1-E-3, Steam Generator Tube Rupture, If RCPs are NOT running, a false red or orange path may be indicated for 1-FR-P.1 during the following steps. T-cold in the ruptured loop should be disregarded until Step 43. Steps 32 to 42 of 1-E-3, stop the safety injection, realign normal charging and letdown and restore normal pressure control. Therefore, a transition to 1-FR-P.1 is not allowed until the SI is terminated in accordance with 1-E-3. It is plausible to believe that a transition to 1-FR-P.1 would be immediately effected as if either a red or orange path is indicated on a status tree, then a transition to that trees restoration procedure is normally mandated.

B. Incorrect: Again, a transition to 1-FR-P.1 is not allowed until after safety injection is terminated in accordance with 1-E-3. It is plausible to believe that a transition would be delayed until after 1-E-3 is completed because one may recall that a restriction on the use of 1-FR-P.1 exists and then misapply such.

C. Incorrect: This distractor is also incorrect and plausible for the same reason as the B distractor.

D. Correct: As described, it is correct that the use of 1-FR-P.1 is not allowed until after SI is terminated in accordance with 1-E-3.

Question Number: 78 Tier: 1 Group: 1 K/A: 038 Steam Generator Tube Rupture 2.4 Emergency Procedures / Plan 2.4.31 Knowledge of annunciator alarms, indications, or response procedures.

Importance Rating: 4.2 4.1 10 CFR Part 55: (CFR: 41.10 / 43.5 / 45.11) 10CFR55.43.b: 10 CFR 55.43(b)(5)

K/A Match: K/A is matched because the applicant is required to correctly implement the procedures 1-E-3 and 1-FR-P.1 during a SGTR.

1-E-3, Steam Generator Tube Rupture Technical

Reference:

TI-12.04, Users Guide for Abnormal and Emergency Operating Instructions Proposed references to None be provided:

Learning Objective: 3-OT-TI1204

24. State the action required when a RED or Orange Path is diagnosed while monitoring the CSF status trees.

Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question for the 2015-301 NRC SRO Exam Comments: See the marked up Clarification Guidance for SRO-only Questions.

WBN Steam Generator Tube Rupture 1-E-3 Unit 1 Rev. 0003 Step Action/Expected Response Response Not Obtained This is the caution immediately prior to the steps directing the initial cooldown to target incore temp.

CAUTION

  • The 1500 psig RCP trip criteria is NOT applicable during or after a controlled RCS cooldown and depressurization.
  • If total feed flow CAPABILITY of 410 gpm is AVAILABLE, 1-FR-H.1, Loss of Secondary Heat Sink, should NOT be implemented.
  • Excessive steam dump cooldown rate will cause MSIV isolation due to the rate sensitive signal.
  • If RCPs are NOT running, a false red or orange path may be indicated for 1-FR-P.1 during the following steps. T-cold in the ruptured loop should be disregarded until Step 43.

If RCPs are not running, do not transition to 1-FR-P.1 until AFTER SI is terminated.

Page 11 of 47

WBN Steam Generator Tube Rupture 1-E-3 Unit 1 Rev. 0003 Step Action/Expected Response Response Not Obtained START THE

18. INITIATE RCS cooldown to INITIAL target Incore temp, determined COOLDOWN from Step 17.
a. DUMP steam to condenser from a. IF condenser steam dumps NOT Intact S/G(s) at maximum available, THEN achievable rate:

USE Intact S/G PORVs at IF dumps are in Tavg mode, maximum achievable cooldown THEN: rate.

1) PLACE steam dump controls OFF. IF an Intact S/G is NOT available,
2) PLACE steam dump mode THEN switch in STEAM PRESSURE. PERFORM one BUT NOT BOTH of the following:
3) ENSURE steam dump demand indicator 1-XI-1-33
  • USE Faulted S/G, reading zero.
4) PLACE steam dump OR controls ON. * ** GO TO 1-ECA-3.1, SGTR
5) PLACE steam dump LOCA - Subcooled Recovery.

controller in MAN, AND FULLY OPEN three cooldown valves

( 25% demand).

Step continued on the next page Page 12 of 47

WBN Steam Generator Tube Rupture 1-E-3 Unit 1 Rev. 0003 Step Action/Expected Response Response Not Obtained CAUTION

  • SI should be terminated as quickly as possible after termination criteria are met to prevent Ruptured S/G overfill.
  • If total feed flow CAPABILITY of 410 gpm is AVAILABLE, 1-FR-H.1, Loss of Secondary Heat Sink, should NOT be implemented.

TERMINATE THE SI.

32. CHECK SI termination criteria:
a. CHECK RCS subcooling greater a. ** GO TO 1-ECA-3.1, SGTR and than 65°F [85°F ADV]. LOCA - Subcooled Recovery.
b. CHECK secondary heat sink b. ** GO TO 1-FR-H.1, Loss of with either: Heat Sink.
  • Total available feed flow greater than 410 gpm, OR
  • At least one S/G NR level greater than 29%

[39% ADV].

c. CHECK RCS pressure c. ** GO TO 1-ECA-3.1, SGTR and stable or rising. LOCA - Subcooled Recovery.
d. CHECK PZR level greater d. ** GO TO Step 16.

than 15% [33% ADV].

STEPS 33 to 42:

stop the safety injection realign normal charging and letdown regain pressure control Page 21 of 47

WBN Steam Generator Tube Rupture 1-E-3 Unit 1 Rev. 0003 Step Action/Expected Response Response Not Obtained

42. (continued)
d. MAINTAIN RCS pressure at d. IF letdown in service, THEN Ruptured S/G pressure:

ALIGN aux spray USING

  • CONTROL PZR heaters Appendix A (1-E-3) as necessary.

ALIGN AUX SPRAY.

  • USE normal PZR spray as necessary.

IF letdown NOT in service, THEN This is the first time (since just before USE one PZR PORV, AND the start of the initial cooldown to MONITOR the following:

target incore

  • Vessel head void formation.

temperature) that a transition to 1-FR-

  • PZR level rise.

P.1 would be

permitted.

NOTE Normal monitoring of T-cold for 1-FR-P.1 can now be resumed.

The Caution prior to Step 18 regarding a false red or orange path is no longer applicable.

43. DETERMINE if Cntmt spray should be stopped:
a. MONITOR Cntmt pressure a. WHEN Cntmt pressure less than 2.0 psig. less than 2.0 psig, THEN PERFORM Substeps 43b thru e.
    • GO TO Step 44.
b. CHECK at least one b. ** GO TO Step 44.

Cntmt spray pump running.

Step continued on next page Page 28 of 47

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]

Some examples of SRO exam items for this topic include:

  • Administrative processes for the installation of temporary instrumentation.
  • Processes for changing the plant or plant procedures.

Section IV provides an example of a satisfactory SRO-only question related to this topic.

D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

[10 CFR 55.43(b)(4)]

Some examples of SRO exam items for this topic include:

  • Process for gaseous/liquid release approvals, i.e., release permits.
  • Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.

Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.

SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.

E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

The question requires the This 10 CFR 55.43 topic involves both 1) assessing plant conditions detailed knowledge (normal, abnormal, or emergency) and then 2) selecting a procedure or of a note in the section of a procedure to mitigate, recover, or with which to proceed. One procedure versus area of SRO level knowledge (with respect to selecting a procedure) is the overall knowledge of the content of the procedure versus knowledge of the strategy. procedures overall mitigative strategy or purpose.

The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:

Page 6 of 16

The question requires the Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) knowledge of the decision point of

  • Knowledge of when to implement attachments and appendices, including whether or not to how to coordinate these items with procedure steps.

transition to 1-FR-

  • Knowledge of diagnostic steps and decision points in the emergency P.1 (e.g. the operating procedures (EOP) that involve transitions to event specific sub-detailed knowledge procedures or emergency contingency procedures.

of the point in the

  • Knowledge of administrative procedures that specify hierarchy, procedure). implementation, and/or coordination of plant normal, abnormal, and emergency procedures.

SRO-only knowledge should not be claimed for questions that can be answered solely using systems knowledge; e.g.:

  • how the system works.
  • system flow path.

Note that the

  • component locations, etc.

question does not require the SRO-only knowledge should not be claimed for questions that can be fundamental answered solely using fundamental knowledge of:

knowledge of the plant parameters

  • the basic purpose, the overall sequence of events that will occur, or the requiring entry into overall mitigative strategy of a procedure.

1-FR-P.1 (i.e.

  • any AOP entry condition.

requiring the RO

  • plant parameters that require direct entry to major EOPs; e.g., major LOK). Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
  • immediate operator actions of a procedure.

Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.

Page 7 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, Yes RO question flowpath, logic, component location?

No Can the question be answered solely by knowing immediate operator actions? Yes RO question No Can the question be answered solely by knowing entry conditions for AOPs or plant parameters Yes RO question that require direct entry to major EOPs?

No Can the question be answered solely by knowing the purpose, overall sequence of events, or Yes RO question overall mitigative strategy of a procedure?

No Does the question require one or more of the following?

  • Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
  • Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps Yes SRO-only
  • Knowledge of diagnostic steps and decision points in the question This bullet could EOPs that involve transitions to event specific sub-also be utilized as procedures or emergency contingency procedures TI-12.04 also
  • Knowledge of administrative procedures that specify delineates the use hierarchy, implementation, and/or coordination of plant of 1-FR-P.1 during normal, abnormal, and emergency procedures Again, the 1-E-3. Question requires No the detailed knowledge of a note (e.g. a Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only decision point -

whether or not to implement 1-FR-P.1).

Page 8 of 16

79.

Given the following timeline:

00:00:00 0-SI-0-3, Weekly Log has been completed and indicates the following TWO items:

ITEM ONE ITEM TWO 00:00:0 1 1-XS-57-96, 125 VITAL BATT BD VOLTMETER SELECTOR is in position I 00: 01:0 1 1-EI-57-96, VIT BATT BDS VOLTS reads 131.5 VDC 00: 1 1 :00 1-EI-57-96 reads 127.5 VDC 00:21:00 1-EI-57-96 reads 123.5 VDC Which ONE of the following describes the FIRST time that LCO 3.8.4, DC Sources -

Operating will NOT be met??

T/S LCO 3.8.4 will FIRST NOT met at time ________.

A. 00:00:00 B. 00:0 1:00 C. 00: 1 1 :00 D. 00:2 1:00

CORRECT ANSWER: A DISTRACTOR ANALYSIS:

A. Correct: As given in the stem of the question, at 00:00:00, data recorded on 0-SI-03 reveals that the CB 2 (the output breaker for 0-CHGR-236-1) is A. This is the nomenclature which stipulates that it is available but not closed or inoperable. Also, breaker 225 (the 125V Vital Batt Bd I breaker which ties a spare battery charger to the board) is available.

Because of these facts, one may ascertain that NO charger is aligned to the Vital Battery Board. The LCO bases for T/S LCO 3.8.4 stipulate, An OPERABLE vital DC electrical power subsystem requires all required batteries and respective chargers to be operating and connected to the associated DC buses. Therefore, because NO charger is aligned to the Vital Batt Bd I and such information was received by the SRO at 00:00:00, actions of T/S LCO 3.8.4 are required at 00:00:00. Additionally, actions of T/S LCO 3.8.4 are necessary through the declaration of SR 3.0.1 which stipulates, Failure to meet the surveillanceshall be failure to meet the LCO.

B. Incorrect: This distractor is incorrect because as mentioned the required actions of the T/S LCO were first required at 00:00:00. It is plausible however because if one did not understand the content of the basis for T/S LCO 3.8.4 and subsequently applied the voltage criteria for Vital Battery V to Vital Battery I, one would arrive at the result that this distractor was correct.

C. Incorrect: This distractor is incorrect because as mentioned the required actions of the T/S LCO were first required at 00:00:00. It is plausible however because if one did not understand the content of the basis for T/S LCO 3.8.4 and subsequently applied the correct voltage criteria for Vital Battery I, one would arrive at the result that this distractor was correct.

D. Incorrect: This distractor is incorrect because as mentioned the required actions of the T/S LCO were first required at 00:00:00. It is plausible however because if one did not understand the content of the basis for T/S LCO 3.8.4 and subsequently applied the voltage criteria for the DG battery to Vital Battery I, one would arrive at the result that this distractor was correct.

Question Number: 79 Tier: 1 Group: 1 K/A: 058 Loss of DC Power 2.4 Emergency Procedures / Plan 2.4.47 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

Importance Rating: 4.2 4.2 10 CFR Part 55: (CFR: 41.10 / 43.5 / 45.12) 10CFR55.43.b: 10 CFR 55.43(b)(2)

K/A Match: K/A is matched because while in a loss of DC power the applicant is required to accurately diagnose the operability of the Vital Battery I.

The applicant must do so using 0-SI-0-3, T/S LCO 3.8.4 and the data trended in a timeline contained in the stem of the question. The question is applicable to the loss of DC power because the loss of a vital charger is an initiator to such casualty.

Technical

Reference:

T/S LCO 3.8.4, DC Sources - Operating T/S LCO 3.8.4 Basis Proposed references to None be provided:

Learning Objective:

Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question for the 2015-301 NRC SRO Exam Comments: See the marked up Clarification Guidance for SRO-only Questions.

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]:

A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)]

Some examples of SRO exam items for this topic include:

  • Reporting requirements when the maximum licensed thermal power output is exceeded.
  • The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
  • National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
  • Processes for TS and FSAR changes.

Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic.

B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

Some examples of SRO exam items for this topic include:

The fact that a charger must be

  • Application of Required Actions (Section 3) and Surveillance aligned for the DC Requirements (SR) (Section 4) in accordance with rules of application source to be requirements (Section 1).

considered

  • Application of generic Limiting Condition for Operation (LCO)

OPERABLE is requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).

contained in the TS

  • Knowledge g of TS bases that are required to analyze TS required actions basis. and terminology.

SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of  1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action statements and the All of the safety limits since Reactor Operators (ROs) are typically required to know distractors rely on these items.

information contained "below SRO-only knowledge generally cannot be claimed for questions that can be the line." answered solely based on expected RO TS knowledge. ROs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Page 3 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)

Note that TS 3.8.4 only specifies that DC sources must be operable. One RO must look in the knowledge basis to determine what a DC source is.

Above this line Page 4 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)

(Tech Specs)

Can question be answered solely by knowing  1 Yes hour TS/TRM Action? RO question No Can question be answered solely by knowing the Yes LCO/TRM information listed above-the-line? RO question No Can question be answered solely by knowing the Yes TS Safety Limits? RO question No Does the question involve one or more of the following for TS, TRM, or ODCM?

  • Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
  • Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)) Yes SRO-only
  • Knowledge g of TS bases that is required to analyze TS question required actions and terminology No Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only Page 5 of 16

DC Sources - Operating 3.8.4 3.8 ELECTRICAL POWER SYSTEMS 3.8.4 DC Sources - Operating LCO 3.8.4 Four channels of vital DC and four Diesel Generator (DG) DC electrical power subsystems shall be OPERABLE.

No mention of the charger is made in the "above the ------------------------------------------------NOTES------------------------------------------------

line" portion of the T/S. 1. Vital Battery V may be substituted for any of the required vital batteries.

2. The C-S DG and its associated DC electrical power subsystem may be substituted for any of the required DGs and their associated DC electrical power subsystem.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One vital DC electrical A.1 Restore vital DC electrical 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> power subsystem power subsystem to inoperable. OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> C. One DG DC electrical C.1 Restore DG DC electrical 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> power subsystem power subsystem to inoperable. OPERABLE status.

(continued)

Watts Bar-Unit 1 3.8-24

DC Sources - Operating 3.8.4 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Declare associated DG Immediately associated Completion inoperable.

Time of Condition C not met.

The S/Rs are the source of the voltage requirements.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.1 Verify vital battery terminal voltage is 128 V (132 V 7 days for vital battery V) on float charge.

SR 3.8.4.2 Verify DG battery terminal voltage is 124 V on float 7 days charge.

SR 3.8.4.3 Verify for the vital batteries that the alternate feeder 7 days breakers to each required battery charger are open.

SR 3.8.4.4 Verify correct breaker alignment and indicated power 7 days availability for each DG 125 V DC distribution panel and associated battery charger.

(continued)

Watts Bar-Unit 1 3.8-25

DC Sources-Operating B 3.8.4 BASES APPLICABLE The OPERABILITY of the DC sources is consistent with the initial assumptions SAFETY ANALYSES of the accident analyses and is based upon meeting the design basis of the (continued) plant. This includes maintaining the DC sources OPERABLE during accident conditions in the event of:

a. An assumed loss of all offsite AC power or all onsite AC power; and
b. A worst case single failure.

The DC sources satisfy Criterion 3 of the NRC Policy Statement.

LCO Four 125V vital DC electrical power subsystems, each vital subsystem channel consisting of a battery bank, associated battery charger and the corresponding control equipment and interconnecting cabling supplying power to the associated Note that a charger DC bus within the channel; and four DG DC electrical power subsystems each must be operating consisting of a battery, a dual battery charger assembly, and the corresponding control equipment and interconnecting cabling are required to be OPERABLE to and connected.

ensure the availability of the required power to shut down the reactor and maintain it in a safe condition after an anticipated operational occurrence (A00) or a postulated DBA. Loss of any DC electrical power subsystem does not prevent the minimum safety function from being performed (Ref. 4).

An OPERABLE vital DC electrical power subsystem requires all required batteries and respective chargers to be operating and connected to the associated DC buses.

The LCO is modified by two Notes. Note 1 indicates that Vital Battery V may be substituted for any of the required vital batteries. However, the fifth battery cannot be declared OPERABLE until it is connected electrically in place of another battery and it has satisfied applicable Surveillance Requirements. Note 2 has been added to indicate that the C-S DG and its associated DC subsystem may be substituted for any of the required DGs. However, the C-S DG and its associated DC subsystem cannot be declared OPERABLE until it is connected electrically in place of another DG, and it has satisfied applicable Surveillance Requirements.

(continued)

Watts Bar-Unit 1 B 3.8-57 Revision 113

80.

Given the following conditions:

- A perturbation occurred on the 161kV transmission grid.

- During the perturbation, the control power supply breaker, for the CSST C load tap changer for the Y winding TRIPS OPEN.

- The following indications are observed on 0-ECB-3:

Which ONE of the following describes the operability of the offsite power supply AND how the CSST C - Y winding voltage will be maintained?

In accordance with T/S LCO 3.8.1, AC Sources - Operating, the offsite power supply described above ____(1)____ operable.

In accordance with 1-PI-OPS-1-500KV, Main Control Room Voltage Monitoring, the associated SDBD will be maintained above the MINIMUM voltage requirement by

____(2)____.

A. (1) IS (2) placing the DG on the SDBD B. (1) IS (2) notifying the Northeast Area Dispatcher (NEAD) to ensure that the 161kV transmission alignments are adequate C. (1) IS NOT (2) placing the DG on the SDBD D. (1) IS NOT (2) notifying the Northeast Area Dispatcher (NEAD) to ensure that the 161kV transmission alignments are adequate

CORRECT ANSWER: B DISTRACTOR ANALYSIS:

A. Incorrect: While it is correct that the offsite power source is currently operable, it is not correct that the emergency diesel generator would be placed on the SDBD to maintain it operable. It is plausible to believe this as if such were done, the voltage of the SDBD would be certainly maintained within limits. Also, it very reasonably seems counterintuitive that one would adjust the entire 161kV grid voltage to compensate for the needs of one generating plant but that is precisely the case. Normally, the load tap changers account for the daily fluctuations in grid voltage. However, upon the loss of a nuclear facilitys capability to adjust for this, dispatch will coordinate with the remaining generating plants to maintain grid voltage.

B. Correct: As seen in 1-PI-OPS-1-500KV, Main Control Room Voltage Monitoring, WHEN CSST tap changer(s) have been placed in any of the following alternative alignments: Common Station Service Transformer C or Load Tap Changer Loss of Power or De-energizedTHEN NOTIFY NEAD of the alternative alignment. The basis for this notification is seen in the note preceding this step: Technical Specification operability is maintained in alternate alignment configuration for CSST Load Tap Changers by ensuring transmission alignments (TRO-TO-SOP-30.130, Watts Bar Nuclear Plant Grid Operating Guide) are adequate to ensure minimum voltage requirements are met. Furthermore, one may refer to 0-SI-82-2, 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Diesel Generator AC power source operability verification to learn that the allowable voltage range for the 6.9kV SDBDs is 6800 to 7260VAC.

Therefore, the offsite power supply remains operable.

C. Incorrect: Again, it is incorrect and yet plausible placing the EDG on the SDBD would ensure that the minimum voltage requirement was met and that the operability of the offsite source was not maintained.

D. Incorrect: While it is correct that a notification would ensure that the minimum voltage require was meet, it is not correct that the operability of the offsite supply was not maintained. It is plausible to believe such because the original design output of the plant required that any time that a tap changer be placed in manual or de-energized, that the affiliated offsite power source be declared inoperable. Additionally, one may believe that 7.08kV is outside of the nominally allowed band (e.g. if they assumed that a +/-100VAC tolerance existed) for the 6.9 kV shutdown board. Notice that the allowable voltage range of 6800-7260 VAC is 100 VAC less than 6900VAC and 360 VAC greater than 6900VAC. Therefore, a +/- 100VAC band would be plausible.

Question Number: 80 Tier: 1 Group: 1 K/A: 077 Generator Voltage and Electric Grid Disturbances AA2. Ability to determine and interpret the following as they apply to Generator Voltage and Electric Grid Disturbances:

AA2.07 Operational status of engineered safety features Importance Rating: 3.6 4.0 10 CFR Part 55: (CFR: 41.5 and 43.5 / 45.5, 45.7, and 45.8) 10CFR55.43.b: 10 CFR 55.43(b)(2)

K/A Match: K/A is matched because the applicant is required to identify the operability of an offsite power source when its compensatory measure for electric grid disturbances is de-energized. The applicant must then identify the correct method by which the minimum voltage requirement would be met.

1-PI-OPS-1-500KV, Main Control Room Voltage Monitoring Technical

Reference:

0-SI-82-2, 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Diesel Generator AC power source operability verification T/S Basis for LCO 3.8.1 Proposed references to None be provided:

Learning Objective: 3-OT-SYS245A 11.DESCRIBE the following aspects of Technical Specifications and Technical Requirements for this system:

a. The conditions and required actions with completion time of one hour or less
b. The Limiting Conditions for Operation, Applicability, and Bases.

Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question for the 2015-301 NRC SRO Exam Comments: See the marked up Clarification Guidance for SRO-only Questions.

WBN Main Control Room 1-PI-OPS-1-500KV Unit 1 Voltage Monitoring Rev. 0007 Page 7 of 13 5.1 Voltage Control Monitoring (continued)

NOTE VARS are to be maintained in accordance with section 5.2.

C. IF 500kV voltage is high, THEN ENSURE Main Generator VARS are incoming.

D. IF 500kV voltage is low, THEN ENSURE Main Generator VARS are outgoing.

NOTE Tap Changers are normally operated in auto but can be operated in manual at SRO discretion. Operation in manual is considered an alternate alignment with respect to the operating requirements and limitations imposed by the WBN grid operating guide.

Technical Specification operability is maintained in alternate alignment configuration for CSST Load Tap Changers by ensuring transmission alignments (TRO-TO-SOP-30.130, Watts Bar Nuclear Plant Grid Operating Guide) are adequate to ensure minimum voltage requirements are met. NEAD shall be notified when the alternate alignments are planned, entered, and exited.

[3] WHEN CSST tap changer(s) have been placed in any of the following alternative alignments:

  • 6.9kV Common Board A or B Loads on Alternate Feeders
  • 480V Turbine Building Common Board A or B on Alternate Feeder
  • Common Station Service Transformer C or D Controls on Alternate Feeder
  • Common Station Service Transformer C or D Load Tap Changer Loss of Power or De-energized
  • Common Station Service Transformer C or D Load Tap Changer in OFF or in Manual During Modes 1 - 4 THEN NOTIFY NEAD of the alternative alignment.

[4] NOTIFY NEAD within 30 minutes when Main Generator Voltage Regulator is NOT in automatic.

End of Section

WBN 8 HOUR DIESEL GENERATOR AC 0-SI-82-2 Unit 0 POWER SOURCE OPERABILITY Rev. 0013 VERIFICATION Page 6 of 31 Date ________ Initials 4.0 PREREQUISITE ACTIONS 4.1 Preliminary Actions

[1] RECORD Start Date and Time on Surveillance Task Sheet. ________

4.2 Approvals and Notifications

[1] OBTAIN SM/SRO approval to perform this instruction on Surveillance Task Sheet. ________

5.0 ACCEPTANCE CRITERIA A. Each qualified offsite power circuit has the correct breaker alignment and indicated power available.

B. Each DG tested is capable of starting from standby condition or modified start and achieving steady state voltage of greater than or equal to 6800 Volts and less than or equal to 7260 Volts and frequency greater than or equal to 58.8 Hz and less than or equal to 61.2 Hz.

AC Sources - Operating B 3.8.1 BASES (continued)

APPLICABLE The initial conditions of DBA and transient analyses in the SAFETY ANALYSES FSAR, Section 6 (Ref. 4) and Section 15 (Ref. 5), assume ESF systems are OPERABLE. The AC electrical power sources are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to ESF systems so that the fuel, Reactor Coolant System (RCS), and containment design limits are not exceeded. These limits are discussed in more detail in the Bases for Section 3.2, Power Distribution Limits; Section 3.4, Reactor Coolant System (RCS); and Section 3.6, Containment Systems.

The OPERABILITY of the AC electrical power sources is consistent with the initial assumptions of the Accident analyses and is based upon meeting the design basis of the plant. This results in maintaining at least two DG's associated with one load group or one offsite circuit OPERABLE during Accident conditions in the event of:

a. An assumed loss of all offsite power or all onsite AC power; and
b. A worst case single failure.

The AC sources satisfy Criterion 3 of NRC Policy Statement.

LCO Two qualified circuits between the Watts Bar Hydro 161 kV switchyard and the onsite Class 1E Electrical Power System and separate and independent DGs for each train ensure availability of the required power to shut down the reactor and One must use this maintain it in a safe shutdown condition after an anticipated operational statement in occurrence (AOO) or a postulated DBA.

addition to the 1-Qualified offsite circuits are those that are described in the FSAR and are part PI-OPS-500kV to ofthe licensing basis for the plant.

determine the operability of the Each offsite circuit must be capable of maintaining acceptable frequency and SDBD. voltage, and accepting required loads during an accident, while connected to the 6.9 kV shutdown boards.

Offsite power from the Watts Bar Hydro 161 kV switchyard to the onsite Class 1E distribution system is from two independent immediate access circuits. Each of the two circuits are routed from the switchyard through a 161 kV transmission line and 161 to 6.9 kV transformer (common station service transformers) to the onsite Class 1E distribution system. The medium voltage power system starts at the low-side of the common station service transformers.

(continued)

Watts Bar-Unit 1 B 3.8-3

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]:

A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)]

Some examples of SRO exam items for this topic include:

  • Reporting requirements when the maximum licensed thermal power output is exceeded.
  • The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
  • National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
  • Processes for TS and FSAR changes.

Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic.

B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

Some examples of SRO exam items for this topic include:

  • Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
  • Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
  • Knowledge of TS bases that are required to analyze TS required actions and terminology.

SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.

SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. ROs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Page 3 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)

RO knowledge Above this line Page 4 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)

(Tech Specs)

Can question be answered solely by knowing 1 Yes hour TS/TRM Action? RO question No Can question be answered solely by knowing the Yes LCO/TRM information listed above-the-line? RO question No The voltage requirements for the SDBDs are contained in the S.R.s Can question be answered solely by knowing the Yes TS Safety Limits? RO question No Does the question involve one or more of the following for TS, TRM, or ODCM?

  • Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
  • Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4) Yes SRO-only
  • Knowledge of TS bases that is required to analyze TS question required actions and terminology No Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only Page 5 of 16

81.

Given the following conditions:

- A LOCA has occurred on Unit 1.

- 1-ECA-1.1, Loss of RHR Sump Recirculation is in progress.

- CNTMT Pressure is RISING .

- The CNTMT Critical Safety Function IS ORANGE.

Which ONE of the following describes the appropriate procedure selection AND operation of the Containment Spray Pumps?

The US will ____(1)____ and direct the crew to operate the Containment Spray Pumps as described in ____(2)____.

NOTE: 1-FR-Z.1, High Containment Pressure 1-ECA-1.1, Loss of RHR Sump Recirculation A. (1) REMAIN in 1-ECA-1.1 (2) 1-ECA-1.1 B. (1) TRANSITION to 1-FR-Z.1 (2) 1-ECA-1.1 C. (1) REMAIN in 1-ECA-1.1 (2) 1-FR-Z.1. High Containment Pressure D. (1) TRANSITION to 1-FR-Z.1 (2) 1-FR-Z.1. High Containment Pressure

CORRECT ANSWER: B DISTRACTOR ANALYSIS:

A. Incorrect: Plausible because there is another instruction (1-ES-1.3) associated with the containment sump that does take precedent over the Orange Path condition (thus a transition would not be made from 1-ES-1.3) and 1-FR-Z.1 does provide for the direct operation of the Containment Spray Pumps.

B. Correct: Correct: The transition to 1-FR-Z.1 is required due to the ORANGE path, but the Containment Spray Pumps are required to be operated in accordance with1- ECA-1.1 as identified in both 1-ECA-1.1 and 1-FR-Z.1.

C. Incorrect: Incorrect: Plausible because there is another instruction (1-ES-1.3) associated with the containment sump that does take precedent over the Orange Path condition (thus a transition would not be made from 1-ES-1.3) and the Containment Spray Pumps are required to be operated in accordance with 1-ECA-1.1 as identified in both 1-ECA-1.1 and 1-FR-Z.1.

D. Incorrect: Plausible because the transition to 1-FR-Z.1 is required due to the ORANGE path, and 1-FR-Z.1 does provide for the direct operation of the Containment Spray Pumps.

Question Number: 81 Tier: 1 Group: 1 K/A: E11 Loss of Emergency Coolant Recirculation EA2. Ability to determine and interpret the following as they apply to the (Loss of Emergency Coolant Recirculation)

EA2.2 Adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments.

Importance Rating: 3.4 4.2 10 CFR Part 55: (CFR: 43.5 / 45.13) 10CFR55.43.b: 10 CFR 55.43(b)(5)

K/A Match: K/A is matched because the question requires interpreting the conditions and adhering to the appropriate conditions within the emergency procedures which are required by the facility's license. SRO because the question requires knowledge of the content of the procedures versus knowledge of the overall mitigation strategy or purpose as well as the assessment of plant conditions, then selecting the procedure with which to proceed.

Technical

Reference:

1-ECA-1.1, Loss of RHR Sump Recirculation 1-FR-Z.1, High Containment Pressure Proposed references to None be provided:

Learning Objective: 3-OT-FRZ0001

2. Discuss the reasons that ECA-1.1, Loss of RHR Sump Recirculation, is given priority over 1-FR-Z.1, High Containment Pressure for directing Containment Spray operation.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: Bank question W/E11 EA2.2 81 which was last used on the 06/2011 WBN NRC exam.

Comments: The question is SRO only as detailed in the distractor analysis. The question meets the general SRO only criteria of Assessment of facility conditions an selection of appropriate procedures during normal, abnormal, and emergency situations.

WBN User's Guide for 0-TI-12.04 Unit 1 & 2 Abnormal and Emergency Rev. 0000 Operating Instructions Page 30 of 57 2.5.2 Conflicts in Rules of Priority (continued)

B. Entry into 1-ECA-0.0, Loss of Shutdown Power, because of the complete loss of both trains of shutdown boards is expected to be a rare occurrence.

1. When 1-ECA-0.0 is implemented, special considerations come into effect.
a. None of the electrically powered safeguards equipment used to restore Critical Safety Functions is operable.
b. None of the FRs can be implemented.
c. A NOTE at the beginning of instruction 1-ECA-0.0 states that Status Trees should be monitored for information only. The FRs should NOT be implemented.
2. Once in 1-ECA-0.0, the operator is NOT allowed to transition to any other instruction until some form of power is restored to the shutdown boards and a transition step is reached.
3. Permission to implement the FRs is NOT granted until some initial status checks and actions are performed by the operator.

C. Certain instructions take precedence over FRs because of their treatment of specific initiating events.

1. Normally, this precedence is identified in a CAUTION or NOTE at the beginning of the specific instruction.
2. 1-ECA-1.1, Loss of RHR Sump Recirculation, directs the operator to perform actions which are intended to conserve RWST level.
a. 1-ECA-1.1 directs the operator to shutdown containment spray pumps based upon containment pressure.
b. This guidance is in conflict with the guidance of 1-FR-Z.1 which directs the operator to maintain all containment spray pumps in service.
c. The guidance of 1-ECA-1.1 takes priority over the guidance of 1-FR-Z.1.
d. 1-FR-Z.1 contains a CAUTION at the beginning of the instruction to remind the operator of this conflict and its correct resolution.

WBN User's Guide for 0-TI-12.04 Unit 1 & 2 Abnormal and Emergency Rev. 0000 Operating Instructions Page 31 of 57 2.5.2 Conflicts in Rules of Priority (continued)

3. 1-ECA-2.1, Uncontrolled Depressurization of All Steam Generators, addresses depressurization, loss of level and resultant feed flow reduction to all steam generators.
a. This condition results in a RED priority on the Heat Sink Status Tree.
b. A CAUTION statement appears at the beginning of 1-ECA-2.1 and 1-FR-H.1 to identify that 1-FR-H.1 should NOT be implemented if the reduced feed flow condition is under the control of the operator.
4. 1-ECA-0.0 addresses a complete loss of shutdown power during which the actions of a FR in effect could NOT be completed successfully.
a. If a complete loss of shutdown power is experienced, transition to 1-ECA-0.0 is required.

The basis for the plausibility for A & b. 1-ECA-0.1 and 1-ECA-0.2 contain a note at the point where normal C. FR implementation can resume. Status Tree conditions should be reevaluated after that point in the instruction.

5. 1-ES-1.3, Transfer to Containment Sump, maintains suction supply to ECCS pumps and injection flowpath to the core.
a. If RWST level reaches the low level setpoint and auto swapover is actuated or required, transition to 1-ES-1.3 is appropriate.
b. 1-ES-1.3 transfer sequence steps are identified by a number on the control board (e.g. #1) to ensure minimum flowpath prior to continuing with the instruction in effect. 1-ES-1.3 should be implemented and completed through the transfer sequence (or transitioned from as directed in 1-ES-1.3).

2.5.3 Termination of EOI Usage A. EOI usage ends in one of the following ways with plant conditions stable:

1. Transition to a normal plant operating instruction, e.g., GOI.
2. On RHR System operation with COLD SHUTDOWN conditions.
3. On RHR System operation with either RHR containment sump recirculation or hot leg recirculation in service and with long term recovery actions being determined by the Technical Support Center.

WBN High Containment Pressure 1-FR-Z.1 Unit 1 Rev. 0001 Step Action/Expected Response Response Not Obtained 3.0 OPERATOR ACTIONS CAUTION If 1-ECA-1.1, Loss of RHR Sump Recirculation, is in effect, the number of Cntmt spray pumps to be operated is directed in 1-ECA-1.1 rather than in this Instruction.

NOTE Adverse containment setpoints [ADV] should be used where provided due to Phase B actuation.

1. ENSURE Cntmt spray operation: ESTABLISH at least one train of Cntmt spray flow.
a. Cntmt spray signal ACTUATED.
b. Cntmt spray pumps RUNNING.
c. Cntmt spray valves 1-FCV-72-2 and 1-FCV-72-39 OPEN.
d. Cntmt spray pump suction valves OPEN:
  • Valves from RWST:

1-FCV-72-21and 1-FCV-72-22 OR

  • Valves from Cntmt sump:

1-FCV-72-44 and 1-FCV-72-45

e. Cntmt spray flow:
  • 1-FI-72-34
  • 1-FI-72-13 Page 3 of 7

WBN Loss of RHR Sump Recirculation 1-ECA-1.1 Unit 1 Rev. 0003 Step Action/Expected Response Response Not Obtained

4. (continued)
c. CHECK number of spray pumps c. STOP and PULL TO LOCK any running equal to number cntmt spray pump NOT required, required. AND CLOSE discharge valve(s) for pump(s) stopped:
  • 1-FCV-72-2 and/or
  • 1-FCV-72-39 MANUALLY OPERATE spray pumps as required.

DO NOT OPERATE cntmt spray pumps as required by FR-Z.1, High Containment Pressure, UNTIL either of the following:

  • Cntmt spray pump suction aligned to cntmt sump, OR
  • RWST makeup sufficient to support cntmt spray pump operation.

WHEN cntmt sump level greater than 28%[36% ADV], THEN PERFORM steps 5, 6, and 7 as necessary.

    • GO TO Step 8 Page 5 of 35

WBN User's Guide for 0-TI-12.04 Unit 1 & 2 Abnormal and Emergency Rev. 0000 Operating Instructions Page 28 of 57 2.4.4 Status Tree Rules of Usage (continued)

C. Status Trees shall be monitored in the following priority:

1. 1-FR-S, Subcriticality,
2. 1-FR-C, Core Cooling,
3. 1-FR-H, Heat Sink,
4. 1-FR-P, PTS,
5. 1-FR-Z, Containment,
6. 1-FR-I, Inventory.

D. If a RED path is diagnosed, then the Function Restoration Instruction will be implemented IMMEDIATELY.

E. If an ORANGE path is diagnosed, then the remaining Status Trees will be checked. If no RED path exits, then the highest priority ORANGE path Function Restoration Instruction will be implemented.

F. Once implemented because of any RED or ORANGE path, that Function Restoration Instruction will be performed to completion or to a point of transition UNLESS a higher priority condition develops.

1. As a Function Restoration Instruction is performed, the status of that tree may change. This change does NOT change the priority of an instruction in progress.
2. If a higher priority condition develops, the instruction in effect should be suspended and the higher priority condition addressed.

G. When no RED or ORANGE path exists, a YELLOW path Function Restoration Instruction can be implemented at the operators discretion.

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]

Some examples of SRO exam items for this topic include:

  • Administrative processes for the installation of temporary instrumentation.
  • Processes for changing the plant or plant procedures.

Section IV provides an example of a satisfactory SRO-only question related to this topic.

D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

[10 CFR 55.43(b)(4)]

Some examples of SRO exam items for this topic include:

  • Process for gaseous/liquid release approvals, i.e., release permits.
  • Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.

Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.

SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.

E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.

The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:

Page 6 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

  • Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
  • Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
  • Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.

SRO-only knowledge should not be claimed for questions that can be answered solely using systems knowledge; e.g.:

  • how the system works.
  • system flow path.
  • component locations, etc.

SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:

  • the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
  • any AOP entry condition.
  • plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
  • immediate operator actions of a procedure.

Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.

Page 7 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, Yes RO question flowpath, logic, component location?

No Can the question be answered solely by knowing immediate operator actions? Yes RO question No Can the question be answered solely by knowing entry conditions for AOPs or plant parameters Yes RO question that require direct entry to major EOPs?

No Can the question be answered solely by knowing the purpose, overall sequence of events, or Yes RO question overall mitigative strategy of a procedure?

No Does the question require one or more of the following?

  • Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
  • Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps Yes SRO-only
  • Knowledge of diagnostic steps and decision points in the question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
  • Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

82.

Given the following conditions:

- Unit 1 is at 100% power.

- 1-SI-99-10-A is in progress.

Subsequently:

- Unit 1 inadvertently trips due to the maintenance activity.

- Control Rod H-4 isstuck at 215 steps withdrawn.

Which ONE of the following describes the appropriate response?

In accordance with ES-0.1, an immediate boration ____(1)____ required.

Assuming that a condition exists where an Immediate Boration IS required, the SRO will assign the responsibility for the performance of 1-AOI-34 to the ____(2)____ in accordance with 0-TI-12.04.

NOTE: 1-SI-99-10-A, 62 Day Functional Test of SSPS Train A and Reactor Trip Breaker A ES-0.1, Reactor Trip Response 1-AOI-34, Immediate Boration 0-TI-12.04, Users Guide for Abnormal and Emergency Operating Instructions A. (1) IS (2) OAC B. (1) IS NOT (2) OAC C. (1) IS (2) BOP/CRO D. (1) IS NOT (2) BOP/CRO

CORRECT ANSWER: D DISTRACTOR ANALYSIS:

A Incorrect: As seen in step 6 of ES-0.1, Reactor Trip Response, the crew is to ENSURE

. all control and shutdown rods fully inserted: RPIs at bottom scale. If a control rod remained withdrawn, the response not obtained would be used. Such response is: IF two or more rods are NOT fully inserted, THEN INITIATE borationREFER TO 1-AOI-34, Immediate Boration. Therefore, it is not correct that an immediate boration would be required on account of one rod which failed to insert. It is plausible to believe such based upon two facts. Firstly, step 6 of ES-0.1 requires that the RNO be utilized whenever any rod is not fully inserted.

Therefore, the very construct of the step would lead one to believe that action would be required on the account of one rod remaining withdrawn; this is due to the fact that step 6 contains two checks: check if all rods are inserted and then, if not, check if two or more rods are not inserted. One could remember the first check and believe that the ultimate action (the boration) depended upon that verification. Secondly, common sense would dictate that a compensatory measure would be required at any time that a reactor trip failed to insert its full negative reactivity (i.e. the failure of a rod to insert).

TI-12.04, Users Guide for Abnormal and Emergency Operating Instructions directs that When running an AOI concurrently with an EOIthe Unit Supervisor/SRO will assign the BOP/CRO operator responsibility for the AOI.

Therefore, it is not correct that the OAC would be assigned the duty of 1-AOI-34.

It is very plausible to believe such as it is the normal duty of the OAC to initiate actions which directly affect the reactivity of the core (e.g. borate and/or dilute).

Therefore, if the SRO had directly entered 1-AOI-34 and not passed such procedure off to an operator, he would direct the actions of such procedure to the OAC. Also, the next several steps of ES-0.1 monitor, control and initiate items which are normally under the responsibility of the BOP/CRO (e.g. the steam generators and secondary plant). Therefore, TI-12.04 requires that the OAC and BOP/CRO perform a role swap in this specific instance.

B. Incorrect: As described, it is correct that the BOP/CRO would be assigned 1-AOI-34.

However it is incorrect and yet plausible 1-AOI-34 would be implemented on account of one rod which failed to insert.

C. Incorrect: While it is correct that 1-AOI-34 would be used whenever two or more rods remained withdrawn post reactor trip, it is not correct and yet plausible for reasons aforementioned that the OAC would perform 1-AOI-34.

D. Correct: It is correct that 1-AOI-34 would be used whenever two or more rods remained withdrawn post reactor trip. Also, it is correct that in accordance with TI-12.04, the BOP/CRO would perform 1-AOI-34.

Question Number: 82 Tier: 1 Group: 2 K/A: 005 Inoperable/Stuck Control Rod 2.4 Emergency Procedures / Plan 2.4.8 Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

Importance Rating: 3.8 4.5 10 CFR Part 55: (CFR: 41.10 / 43.5 / 45.13) 10CFR55.43.b: 10 CFR 55.43(b)(5)

K/A Match: K/A is matched because the applicant is required to understand when and how 1-AOI-34, Immediate Boration is used in conjunction with ES-0.1, Reactor Trip Response. Such use occurs during the failure of control rod(s) to insert post reactor trip.

Technical 1-SI-99-10-A, 62 Day Functional Test of SSPS Train A and Reactor Trip Breaker A

Reference:

ES-0.1, Reactor Trip Response 1-AOI-34, Immediate Boration TI-12.04, Users Guide for Abnormal and Emergency Operating Instructions Proposed references to None be provided:

Learning Objective: 3-OT-AOI3400

8. DESCRIBE the reasons for the following responses as they apply to 1-AOI-34, Immediate Boration and the following:

When emergency boration is required Actions contained in EOP for emergency boration Cognitive Level:

Higher Lower X Question Source:

New X Modified Bank Bank Question History: New question for the 2015-301 NRC SRO Exam Comments: See the marked up Clarification Guidance for SRO-only Questions.

WBN Reactor Trip Response ES-0.1 Unit 1 Rev. 0024 Step Action/Expected Response Response Not Obtained

6. ENSURE all control and shutdown IF two or more rods are NOT fully rods fully inserted: inserted, THEN
  • RPIs at bottom scale.

INITIATE boration of 3250 gals of greater than or equal to 6120 ppm boron for each rod not fully inserted:

  • REFER TO AOI-34, Immediate Boration.
7. ANNOUNCE reactor trip over PA system.
8. MONITOR S/G levels:
a. At least one S/G NR level greater a. ENSURE feed flow greater than than 29%. 410 gpm.
b. S/G NR levels less than 50% and b. IF any S/G NR level continues to controlled. rise, THEN ISOLATE feed flow to affected S/G.
9. CONTROL S/G NR levels between 29% and 50%.
10. INITIATE BOP realignment:

Page 7 of 21

WBN User's Guide for 0-TI-12.04 Unit 1 & 2 Abnormal and Emergency Rev. 0000 Operating Instructions Page 35 of 57 2.7 Prudent Operator Actions (continued)

3. The operator should consult nearby personnel who are suitably qualified and notify them of their proposed actions. If no disagreement is forthcoming, he should then take the necessary mitigation or preemptive actions to terminate the event.
4. The STAR principle should be applied --Stop, Think, Act, Review. Ask yourself: If I take this action, could I inadvertently cause other more severe problems? Am I better off taking no action at all? How will safety status be affected?

2.8 Use of AOIs While in EOIs

1. During performance of the 1-ES-0.1, if plant conditions warrant implementation of an AOI, then the required AOI may be performed concurrently (on a not-to-interfere basis) with the EOIs.
2. When running an AOI concurrently with an EOI (1-ECA-0.0, 1-ES-0.1, etc.)

the Unit Supervisor/SRO will assign the BOP/CRO operator responsibility for the AOI if another Unit Supervisor is NOT available. If the BOP/CRO operator performs an AOI, he/she should consult directly with the Unit Supervisor and give them the status as required by the AOI.

3. When an AOI in effect directs a Reactor Trip, then the performance of the AOI should continue immediately following transition to 1-ES-0.1.

Performance assignments will be at the discretion of the SM/US based on the status and importance of events in progress.

4. When implementing an AOI outside the horseshoe in the control room, the Unit Supervisor should accompany the board operator to read the procedure steps and direct actions of the operator, unless higher priority conditions demanding the Unit Supervisors attention exist; in which case the BOP/CRO should implement the AOI using the single performer method. The actively licensed STA may serve as a reader unless the crew is in progress of performing actions within the EOI network.

3.0 RECORDS None

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]

Some examples of SRO exam items for this topic include:

  • Administrative processes for the installation of temporary instrumentation.
  • Processes for changing the plant or plant procedures.

Section IV provides an example of a satisfactory SRO-only question related to this topic.

D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

[10 CFR 55.43(b)(4)]

Some examples of SRO exam items for this topic include:

  • Process for gaseous/liquid release approvals, i.e., release permits.
  • Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.

Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.

SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.

E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.

The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:

Page 6 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

  • Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
  • Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
  • Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.

SRO-only knowledge should not be claimed for questions that can be answered solely using systems knowledge; e.g.:

  • how the system works.
  • system flow path.
  • component locations, etc.

SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:

  • the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
  • any AOP entry condition.
  • plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
  • immediate operator actions of a procedure.

Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.

Page 7 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, Yes RO question flowpath, logic, component location?

No Can the question be answered solely by knowing immediate operator actions? Yes RO question No Can the question be answered solely by knowing entry conditions for AOPs or plant parameters Yes RO question that require direct entry to major EOPs?

No Can the question be answered solely by knowing the purpose, overall sequence of events, or Yes RO question overall mitigative strategy of a procedure?

No Does the question require one or more of the following?

  • Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
  • Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps Yes SRO-only
  • Knowledge of diagnostic steps and decision points in the question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
  • Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

83.

Given the following conditions:

- 1-AOI-34, Immediate Boration, Section 3.2, Boration of RCS with CVCS in Service is in progress.

- 1-HS-62-140A, VCT MAKEUP CONTROL has JUST been taken to START and released.

Subsequently:

- 1-TANK-62-239, Boric Acid Tank A outlet piping RUPTURES.

- 1-FI-62-139, BA TO BLENDER FLOW is 0 gpm.

- 1-LI-62-238, BAT A LEVEL is RAPIDLY LOWERING .

Which ONE of the following completes the statement below?

In order to borate the RCS, the US must __________.

A. place the C BAT in service to the U1 CVCS blender in accordance with 1-SOI-62.05 B. continue in section 3.2 of 1-AOI-34 and align the RWST to the charging pump suction C. continue in section 3.2 of 1-AOI-34 and place the C BAT in service to the U1 CVCS blender D. alternate the charging pump suction to and from the RWST Using 1-LCV-62-135 and 136 in accordance with 1-SOI-62.01

CORRECT ANSWER: B DISTRACTOR ANALYSIS:

A. Incorrect: 1-AOI-34 will sequentially attempt methods of boration. Section 3.2 of such procedure will eventually direct the crew to place the Charging Pumps suction on the RWST. Therefore, 1-AOI-34 would provide a means of borating the RCS and because of this, the use of 1-SOI-62.05 is not required. This distractor is plausible because physically, it would be effective at providing a means of borating the RCS.

B. Correct: Again, 1-AOI-34 would permit the alignment of the Charging Pump suctions to the RWST such that the RCS could be borated.

C Incorrect: 1-AOI-34 does not alternate which BAT is in service to the boric acid transfer pumps.

. Additionally, it does not make any provision for placing an additional BAT in service. It is plausible to believe that it made such because that would be a very reasonable provision to be had in the event of a mechanical failure or low tank level. Additionally, if a very large boration were required (such as that required if multiple rods were stuck out after a reactor trip), the procedure does not directly provide for either BAT makeup or BAT swap (there are checks to validate that level is above that required by the T/Rs and subsequent REFER steps which direct the crew to an SOI).

D Incorrect: The SOIs would not be used to provide an Immediate Boration given the conditions in the

. stem of this question. This particular distractor is plausible as it would provide a means of boration which is equivalent to that yielded by the AOI. Again, it is incorrect because the US is not required to utilize it (as 1-AOI-34 does provide a means of boration).

Question Number: 83 Tier: 1 Group: 2 K/A: 024 Emergency Boration AA2. Ability to determine and interpret the following as they apply to the Emergency Boration:

AA2.04 Availability of BWST Importance Rating: 3.4 4.2 10 CFR Part 55: (CFR: 43.5 / 45.13) 10CFR55.43.b: 10 CFR 55.43(b)(5)

K/A Match: K/A is matched because the applicant must understand the impact of losing the Boric Acid Tank on an emergency boration and then correctly use the Emergency Boration procedure (1-AOI-34).

Technical

Reference:

1-AOI-34, Immediate Boration 0-SOI-62.05, Boric Acid Batching, Transfer, And Storage 1-SOI-62.01, CVCS-Charging and Letdown Proposed references to None be provided:

Learning Objective: 3-OT-AOI3400

9. DETERMINE the following as they apply to 1-AOI-34, Immediate Boration Availability of Boric Acid Tanks Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question for the 2015-301 NRC SRO Exam Comments: The question is SRO only as detailed in the distractor analysis. The question meets the general SRO only criteria of Assessment of facility conditions an selection of appropriate procedures during normal, abnormal, and emergency situations.

WBN Immediate Boration 1-AOI-34 Unit 1 Rev. 0001 3.0 OPERATOR ACTIONS The question 3.1 Diagnostics places the applicant in this section.

IF GO TO Subsection Page CVCS in service to RCS 3.2 5 CVCS shutdown or boration is required during 3.3 14 Refueling Page 4 of 24

WBN Immediate Boration 1-AOI-34 Unit 1 Rev. 0001 Step Action/Expected Response Response Not Obtained 3.2 Boration of RCS with CVCS in Service NOTE Boric acid addition should be noted to assist in determination of reactivity changes.

1. INITIATE normal boration to change CB as necessary:
a. PLACE 1-HS-62-140B MODE SELECTOR, to BOR.
b. CHECK 1-FC-62-139, BA TO BLENDER, indicates GPM.
c. ADJUST 1-FC-62-139, BA TO BLENDER, setpoint to desired The question stem flow rate. places the applicant at this
d. ADJUST 1-FQ-62-139 BA step.

BATCH COUNTER, to ensure boration continues.

e. () MOMENTARILY PLACE 1-HS-62-140A, VCT MAKEUP CONTROL, to START and RELEASE. This is 0 due to the rupture.
f. CHECK 1-HS-62-140A, Red light LIT.
g. VERIFY boric acid flow indication on 1-FI-62-139, BA TO BLENDER FLOW.
2. ENSURE PW to blender isol, 1-FCV-62-143, CLOSED.
3. CHECK PW to blender flow, DISPATCH Operator to CLOSE PW to 1-FI-62-142, indicating ZERO. blender isolation, 1-ISV-62-933

[A4V/713].

Page 5 of 24

WBN Immediate Boration 1-AOI-34 Unit 1 Rev. 0001 Step Action/Expected Response Response Not Obtained 3.2 Boration of RCS with CVCS in Service (continued)

NOTE A delay of 15 to 20 minutes may be expected before effects of negative reactivity insertion are observed.

Normal Boration will NOT be as it is going to the floor of

4. MONITOR for negative reactivity IF normal boration NOT insertingthe aux building.

insertion: negative reactivity, THEN

  • Neutron flux dropping.
    • GO TO Step 6.
  • Tavg dropping.
5. IF normal boration effective, THEN This is impossible
6. ESTABLISH required emergency required to be boration flow: aligned (the "A" BAT is still ruptured).
a. PLACE both BA pumps in FAST speed.
b. () ADJUST emergency borate b. () Locally ADJUST valve 1-FCV-62-138 to obtain 1-FCV-62-138 to obtain required required flow. flow.
c. CHECK emergency borate flow c. () Locally OPEN manual boration on 1-FI-62-137A. valve, 1-ISV-62-929 [Blender Station/713].

ENSURE BA flow control, 1-FCV-62-140, OPEN.

ENSURE BA to Blender, 1-FI-62-139, indicating flow.

Page 6 of 24

WBN Immediate Boration 1-AOI-34 Unit 1 Rev. 0001 Step Action/Expected Response Response Not Obtained 3.2 Boration of RCS with CVCS in Service (continued)

7. IF emergency boration flow established, THEN
    • GO TO Step 9. This will provide the success path
8. ALIGN RWST to CCP suction: for borating the RCS.
a. () OPEN RWST outlet valves 1-LCV-62-135 and 1-LCV-62-136. [C.1]
b. CLOSE VCT outlet valves 1-LCV-62-132 and 1-LCV-62-133.
9. REFER TO the following tech Specs:

Tavg > 200°F.

Tavg 200°F.

  • 3.1.6, Shutdown Bank Insertion Limits.
  • 3.1.7, Control Bank Insertion Limits.
  • 3.4.2, RCS Minimum Temperature for Criticality.
  • 3.5.4, Refueling Water Storage Tank (RWST).
  • 3.9.1, Boron Concentration.

Page 7 of 24

WBN Boric Acid 0-SOI-62.05 Unit 0 Batching, Transfer, And Storage Rev. 0001 Page 41 of 124 Date________ Initials 8.2 Alternate BAT Operation This alignment is physically possible.

8.2.1 Place BAT C In Service With BA Pumps 1A & 1B Aligned to However, if the BATs were BAT C being credited for the Boration flow source/flow path, then a T/R impact would be had.

NOTE This Sect places BAT & BA Pumps in an alternate configuration, inconsistent with 1-TRI-62-3.

[1] MOMENTARILY PLACE 1-HS-62-140A, VCT MAKEUP CONTROL [1-M-6], to STOP and RELEASE. ________

[2] CHECK 1-HS-62-140A Green light LIT. ________

[3] PERFORM the following:

NOMENCLATURE LOCATION POSITION UNID PERF INITIAL BA PMP A 1-M-6 STOP 1-HS-62-230A BA PMP B 1-M-6 STOP 1-HS-62-232A

[4] OPEN 1-ISV-62-1053B, BA XFER PUMP 1B-B DISCHARGE

[A12R/713]. ________

[5] CLOSE the following valves:

NOMENCLATURE LOCATION UNID PERF INITIAL BA XFER PUMP 1A-A A12R/713 1-ISV-62-1054A RECIRC ISOL BORIC ACID TANK A A12R/713 1-ISV-62-1049 OUTLET

[6] OPEN 1-ISV-62-1048A, BA PUMP 1A-A/1B-B CROSSTIE

[A12R/713]. ________

WBN Boric Acid 0-SOI-62.05 Unit 0 Batching, Transfer, And Storage Rev. 0001 Page 42 of 124 Date________ Initials 8.2.1 Place BAT C In Service With BA Pumps 1A & 1B Aligned to BAT C (continued)

[7] IF Boric Acid Filter is bypassed, THEN ENSURE the following:

A. CLOSE 1-ISV-62-1055A, BA XFER PUMP 1A-A BA FLTR A BYPASS. ________

B. OPEN 1-ISV-62-1055B, BA XFER PUMP 1B-B BA FLTR B BYPASS. ________

NOTE Both U1 BA Pumps are now aligned to BAT C.

[8] START desired Boric Acid Pump (N/A pump NOT started):

NOMENCLATURE LOCATION POSITION UNID PERF INITIAL BA PMP A 1-M-6 START 1-HS-62-230A BA PMP B 1-M-6 START 1-HS-62-232A

[9] MOMENTARILY PLACE 1-HS-62-140A, VCT MAKEUP CONTROL [1-M-6], to START and RELEASE. ________

[10] CHECK 1-HS-62-140A, Red light LIT. ________

[11] WHEN desired to return to NORMAL alignment, THEN PERFORM the following:

NOMENCLATURE LOCATION POSITION UNID PERF INITIAL VCT MAKEUP 1-M-6 STOP 1-HS-62-140A CONTROL BA PMP A 1-M-6 STOP 1-HS-62-230A BA PMP B 1-M-6 STOP 1-HS-62-232A

WBN Boric Acid 0-SOI-62.05 Unit 0 Batching, Transfer, And Storage Rev. 0001 Page 43 of 124 Date________ Initials 8.2.1 Place BAT C In Service With BA Pumps 1A & 1B Aligned to BAT C (continued)

[12] PERFORM the following:

NOMENCLATURE LOCATION POSITION UNID PERF VERIF INITIAL INITIAL BA XFER PUMP 1B-B A12R/713 CLOSED 1-ISV-62-1053B DISCHARGE IV BA XFER PUMP 1A-A A12R/713 OPEN 1-ISV-62-1054A RECIRC ISOL IV BORIC ACID TANK A A12R/713 OPEN 1-ISV-62-1049 OUTLET IV BA PUMP 1A-A/1B-B A12R/713 CLOSED 1-ISV-62-1048A CROSSTIE IV

[13] IF Boric Acid Filter is bypassed, THEN ENSURE the following:

A. CLOSE 1-ISV-62-1055B, BA XFER PUMP 1B-B BA FLTR B BYPASS. ________

B. OPEN 1-ISV-62-1055A, BA XFER PUMP 1A-A BA FLTR A BYPASS.

[14] START BA Pump 1A using 1-HS-62-230A, BA PMP A [1-M-6]. ________

[15] MOMENTARILY PLACE 1-HS-62-140A, VCT MAKEUP CONTROL [1-M-6], to START and RELEASE. ________

[16] CHECK 1-HS-62-140A, Red light LIT. ________

[17] IF BAT C is required to be in service, THEN PERFORM Sect 6.5, BAT C Normal Alignment. ________

End of Section

WBN CVCS-Charging and Letdown 1-SOI-62.01 Unit 1 Rev. 0000 Page 82 of 108 Date ________ INITIALS 8.14 Alternating Charging Pump Suction To and From the RWST Using 1-LCV-62-135 And 136

[1] IF transferring Charging Pump Suction from the VCT to the RWST is desired, THEN PERFORM the following:

[1.1] OPEN RWST to CVCS Charging Pump suction:

PERF NOMENCLATURE LOCATION UNID INITIAL RWST TO CHARGING 1-M-5 1-HS-62-135A PMPS SUCTION RWST TO CHARGING 1-M-5 1-HS-62-136A PMPS SUCTION

[1.2] CLOSE the following:

PERF NOMENCLATURE LOCATION UNID INITIAL VCT TO CHARGING 1-M-5 1-HS-62-132A PMPS SUCTION VCT TO CHARGING 1-M-5 1-HS-62-133A PMPS SUCTION

[1.3] ENSURE 1-FCV-62-1228 and 1-FCV-62-1229, CCP SUCTION TO VCT VENT HDR ISOL, CLOSED (green lights LIT). ________

WBN CVCS-Charging and Letdown 1-SOI-62.01 Unit 1 Rev. 0000 Page 83 of 108 Date ________ INITIALS 8.14 Alternating Charging Pump Suction To and From the RWST Using 1-LCV-62-135 And 136 (continued)

[2] IF transferring Charging Pump Suction from the RWST to the VCT is desired, THEN PERFORM the following:

[2.1] ENSURE the following to align VCT to charging pumps:

PERF VERIFIER NOMENCLATURE LOCATION POSITION UNID INITIAL INITIAL VCT TO CHARGING 1-M-5 OPEN 1-HS-62-132A PMPS SUCTION A-P AUTO IV VCT TO CHARGING 1-M-5 OPEN 1-HS-62-133A PMPS SUCTION A-P AUTO IV

[2.2] ENSURE 1-FCV-62-1228 and 1-FCV-62-1229, CCP SUCTION TO VCT VENT HDR ISOL, OPEN (red lights LIT). ________

[2.3] ENSURE RWST to CVCS Charging Pump alignment:

PERF VERIFIER NOMENCLATURE LOCATION POSITION UNID INITIAL INITIAL RWST TO CHARGING 1-M-5 CLOSED 1-HS-62-135A PMPS SUCTION A-P AUTO IV RWST TO CHARGING 1-M-5 CLOSED 1-HS-62-136A PMPS SUCTION A-P AUTO IV End of Section

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]

Some examples of SRO exam items for this topic include:

  • Administrative processes for the installation of temporary instrumentation.
  • Processes for changing the plant or plant procedures.

Section IV provides an example of a satisfactory SRO-only question related to this topic.

D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

[10 CFR 55.43(b)(4)]

Some examples of SRO exam items for this topic include:

  • Process for gaseous/liquid release approvals, i.e., release permits.
  • Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.

Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.

SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.

E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.

The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:

Page 6 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

  • Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
  • Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
  • Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.

SRO-only knowledge should not be claimed for questions that can be answered solely using systems knowledge; e.g.:

  • how the system works.
  • system flow path.
  • component locations, etc.

SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:

  • the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
  • any AOP entry condition.
  • plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
  • immediate operator actions of a procedure.

Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.

Page 7 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, Yes RO question flowpath, logic, component location?

No The question cannot be answered solely on systems Can the question be answered solely by knowing knowledge because such immediate operator actions? Yes RO question knowledge would cause the No applicant to simultaneously arrive at two answers (because knowledge of the Can the question be answered solely by knowing procedures is required).

entry conditions for AOPs or plant parameters Yes RO question that require direct entry to major EOPs?

No Can the question be answered solely by knowing the purpose, overall sequence of events, or Yes RO question overall mitigative strategy of a procedure?

No The question is asking details of Does the question require one or more of the following? the procedure beyond the overall

  • Assessing plant conditions (normal, abnormal, or strategy.

emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed

  • Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps Yes SRO-only
  • Knowledge of diagnostic steps and decision points in the question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
  • Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant The question is normal, abnormal, and emergency procedures asking whether one No must select an SOI to provide a success path or Question might not be linked to proceed with the 10 CFR 55.43(b)(5) for SRO-only AOI in use.

Page 8 of 16

84.

Given the following timeline:

00:00:00 Unit 1 is in The shutdown rods are FULLY WITHDRAWN.

00:0 1 :00 Source Range Nuclear Instrument N-31 fails LOW.

The crew implements 1-AOI-4, Nuclear Instrumentation Malfunctions.

00: 10 : 00 Source Range Nuclear Instrument N-32 fails LOW.

The OAC takes 1-RT-1, REACTOR TRIP to TRIP.

The US enters T/S LCO 3.3.1 condition L:

Which ONE of the following completes the statements below?

In accordance with 0-TI-12.04, 1-E-0 ____(1)____ be entered to confirm the reactor trip.

In accordance with the Unit 1 T/S, SR 3.1.1.1 MUST be completed by _____(2)_____.

NOTE: 1-E-0, Reactor Trip or Safety Injection T/S LCO, 3.3.1 Reactor Trip System (RTS) Instrumentation 0-TI-12.04, Users Guide for Abnormal and Emergency Operating Instructions A. (1) MUST (2) 0 1: 10 : 00 B. (1) MUST (2) 0 1:25:00 C. (1) NEED NOT (2) 0 1: 10 : 00 D. (1) NEED NOT (2) 0 1:25:00

CORRECT ANSWER: C DISTRACTOR ANALYSIS:

A. Incorrect: 0-TI-12.04, demonstrates that the EOI network contains implementation points which are applicable in Modes 1,2,3 or 4.

Therefore, it is incorrect that with the Unit in Mode 5, that the EOI network would be implemented. It is plausible to believe this because there is no procedure other than 1-E-0 which is written to address a reactor trip response and that the stem of the question presents the applicant with a reactor trip.

SR 3.0.2 states that If a Completion Time requires periodic performance on a "once per . . ." basis, the above Frequency extension applies to each performance after the initial performance.

Therefore, the first performance of SR 3.1.1.1 is required 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after the T/S LCO required action entry time or at 0110.

B. Incorrect: Again it is incorrect and yet plausible that the Reactor Trip Response procedure would be used following a reactor trip in Mode 5. It is also incorrect that the first performance of SR 3.1.1.1 would be required at 0125. It is plausible to believe this as SR 3.0.2 does provide that:

The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency.

However, as mentioned, this is only applicable to performances conducted after the initial.

C. Correct: It is correct that 1-E-0 need not be entered in Mode 5 to confirm a reactor trip. It is also correct that the first performance of the surveillance is required at 0110.

D. Incorrect: While it is correct that 1-E-0 need not be entered in Mode 5 to confirm a reactor trip, it is incorrect and yet plausible that the first performance of the surveillance is required at 0125.

Question Number: 84 Tier: 1 Group: 2 K/A: 032 Loss of Source Range Nuclear Instrumentation AA2. Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation:

AA2.06 Confirmation of reactor trip Importance Rating: 3.9 4.1 10 CFR Part 55: (CFR: 43.5 / 45.13) 10CFR55.43.b: 10 CFR 55.43(b)(2) and 10 CFR 55.43(b)(5)

K/A Match: K/A is matched because while injected into a loss of source range instrumentation, the applicant is required to determine if 1-E-0 is required to confirm a reactor trip in Mode 5. The second part of the question requires the applicant to determine when the first performance of a shutdown margin verification would be required post reactor trip.

Technical

Reference:

T/S LCO 3.3.1, Reactor Trip System (RTS) Instrumentation 0-TI-12.04, Users Guide for Abnormal and Emergency Operating Instructions Proposed references to None be provided:

Learning Objective: 3-OT-SYS092A

15. Given a set of plant conditions/parameters, APPLY the appropriate Technical Specifications and Technical Requirements.

3-OT-EOP0000

8. Analyze a set of plant conditions and identify required procedure transitions
15. Explain the purpose for and the basis of each step in E-0 Cognitive Level:

Higher Lower X Question Source:

New X Modified Bank Bank Question History: New question for the 2015-301 NRC SRO Exam Comments:

WBN User's Guide for 0-TI-12.04 Unit 1 & 2 Abnormal and Emergency Rev. 0000 Operating Instructions Page 7 of 57 2.1.2 Mode Applicability of the EOIs The EOIs are written to mitigate emergency transients initiated when the unit is at hot or power conditions.

A. The guidance for operator action in the EOIs assumes that the safety-related equipment required by Tech Specs in Mode 1 or Mode 2 is available for use.

B. The operating crew should implement the EOI network whenever reactor trip or safety injection events are initiated with the unit in Modes 1, 2, or 3.

C. The operating crew should implement the EOI network for the complete loss of shutdown power event with the unit in Modes 1, 2, 3, or 4.

D. Implementation of the EOI network in Mode 4 requires the operating crew to consider plant conditions and each specific instructions applicability.

The EOIs are

1. The EOI network assumes that the Residual Heat Removal (RHR) system never REQUIRED is aligned for its Emergency Core Cooling mode.

to be used in MODE 5. 2. Although most of the FRs can be utilized to respond to events during Mode 4 conditions, they assume ECCS equipment has operated and steam generators are available and required for heat removal.

3. Events (other than complete loss of shutdown power) initiated with the unit in Modes 4, 5, or 6 should be mitigated by implementation of the Abnormal Operating Instructions (AOIs).
4. The operating team should consider implementation of the EOI network if events initiated in Modes 4 or 5 result in plant heat-up to Mode 3.
5. A specific task (e.g., alignment of RHR sump recirc to SI pump suction) that is detailed in the EOIs may be appropriate during an unanticipated event. When considering such actions, the crew must be cautious and NOT apply the instruction out of context.

Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-2 (continued)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify flow is within limits. Once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after 25% RTP AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter The first performance of a Example 1.4-2 has two Frequencies. The first is a one time performance T/S required SR Frequency, and the second is of the type shown in Example 1.4-1. The logical (with a thereafter connector "AND" indicates that both Frequency requirements must be met. Each frequency) does time reactor power is increased from a power level < 25% RTP to 25% RTP, not receive the the Surveillance must be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

alleviation of SR The use of "once" indicates a single performance will satisfy the specified 3.0.2. Frequency (assuming no other Frequencies are connected by "AND"). This type of Frequency does not qualify for the extension allowed by SR 3.0.2. "Thereafter" indicates future performances must be established per SR 3.0.2, but only after a specified condition is first met (i.e., the "once" performance in this example). If reactor power decreases to < 25% RTP, the measurement of both intervals stops. New intervals start upon reactor power reaching 25% RTP.

(continued)

Watts Bar-Unit 1 1.4-3

SR Applicability 3.0 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3.

Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.

SR 3.0.2 The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.

For Frequencies specified as "once," the above interval extension does not apply.

If a Completion Time requires periodic performance on a "once per . . ." basis, the above Frequency extension applies to each performance after the initial performance.

Exceptions to this Specification are stated in the individual Specifications.

SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the risk impact shall be managed.

If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.

(continued)

Watts Bar-Unit 1 3.0-4 Amendment 42

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]:

A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)]

Some examples of SRO exam items for this topic include:

  • Reporting requirements when the maximum licensed thermal power output is exceeded.
  • The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
  • National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
  • Processes for TS and FSAR changes.

Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic.

B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

Some examples of SRO exam items for this topic include:

  • Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
  • Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
  • Knowledge of TS bases that are required to analyze TS required actions and terminology.

SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.

SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. ROs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Page 3 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)

RO knowledge Above this line Page 4 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)

(Tech Specs)

Can question be answered solely by knowing 1 Yes hour TS/TRM Action? RO question No Can question be answered solely by knowing the Yes LCO/TRM information listed above-the-line? RO question No Can question be answered solely by knowing the Yes TS Safety Limits? RO question No Does the question involve one or more of the following for TS, TRM, or ODCM?

  • Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
  • Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4) Yes SRO-only
  • Knowledge of TS bases that is required to analyze TS question required actions and terminology The question No requires the knowledge of SR 3.0.2.

Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only Page 5 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]

Some examples of SRO exam items for this topic include:

  • Administrative processes for the installation of temporary instrumentation.
  • Processes for changing the plant or plant procedures.

Section IV provides an example of a satisfactory SRO-only question related to this topic.

D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

[10 CFR 55.43(b)(4)]

Some examples of SRO exam items for this topic include:

  • Process for gaseous/liquid release approvals, i.e., release permits.
  • Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.

Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.

SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.

E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.

The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:

Page 6 of 16

The question Clarification Guidance for SRO-only Questions requires the Rev 1 (03/11/2010) knowledge of the

  • Knowledge of when to implement attachments and appendices, includingMODES of how to coordinate these items with procedure steps. applicability for the
  • Knowledge of diagnostic steps and decision points in the emergency EOP set (i.e. the operating procedures (EOP) that involve transitions to event specific sub-knowledge of procedures or emergency contingency procedures. which modes
  • Knowledge of administrative procedures that specify hierarchy, require the use of implementation, and/or coordination of plant normal, abnormal, and the EOPs and emergency procedures. which require the use of the AOIs).

SRO-only knowledge should not be claimed for questions that can be answered solely using systems knowledge; e.g.:

  • how the system works.
  • system flow path.
  • component locations, etc.

SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:

  • the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
  • any AOP entry condition.
  • plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
  • immediate operator actions of a procedure.

Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.

Page 7 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, Yes RO question flowpath, logic, component location?

No Can the question be answered solely by knowing immediate operator actions? Yes RO question No Can the question be answered solely by knowing entry conditions for AOPs or plant parameters Yes RO question that require direct entry to major EOPs?

No Can the question be answered solely by knowing the purpose, overall sequence of events, or Yes RO question overall mitigative strategy of a procedure?

No Does the question require one or more of the following?

  • Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
  • Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps Yes SRO-only
  • Knowledge of diagnostic steps and decision points in the question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
  • Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

85.

Given the following conditions:

- The MAXIMUM observed Containment pressure was 0.1 psig.

- An overpressure condition exists on the #1 SG.

- The operating crew has completed steps 1 and 2 of FR-H.2.

- The crew CHECKS affected SG NR level and notes the following:

1-M-4 Indications ICS indication In accordance with FR-H.2 and the associated basis document, which ONE of the following identifies the crew response?

The crew will identify that __________.

NOTE: FR-H.2, Steam Generator Overpressure FR-H.3, Steam Generator High Level A. neither level shown above indicates that the SG is potentially filled solid with water and as such, the crew will REMAIN in FR-H.2 B. ONLY the level shown on ICS indicates that the SG is potentially filled solid with water but because ONLY the 1-M-4 indications may used for a decision point, the crew will REMAIN in FR-H.2 C. BOTH the levels shown on ICS and 1-M-4 indicate that the SG is potentially filled solid with water; the crew will TRANSITION to FR-H.3 D. ONLY the level shown on ICS indicates that the SG is potentially filled solid with water and because the ICS data may be used for a decision point, the crew will TRANSITION to FR-H.3

CORRECT ANSWER: B DISTRACTOR ANALYSIS:

A. Incorrect: As seen in the Westinghouse basis document for FR-H.2, Steam Generator Overpressure, If the level is greater than [93%], the SG water level may be above the narrow range or the SG may even be filled solid with water. Therefore, the value observed on ICS (the computer screen) is greater than 93% and as such represents that the SG may be filled with water. Therefore, it is incorrect to believe that neither level shown indicates that the S/G is potentially filled solid.

B. Correct: As described, the 94% NR SG level observed on ICS indicates that the SG is potentially filled solid with water. As seen in TI-12.04, User's Guide for Abnormal and Emergency Operating Instructions, During performance of the EOI set, the operator is required to utilize PAM instruments when they are provided on the control board.

Therefore, as the 1-M-4 indications are all 92% and therefore, less than the value requiring a transition to FR-H.3, the crew will remain in FR-H.2.

C. Incorrect: Again, only the ICS value is greater than 93%; therefore, it is incorrect to believe that all indications relate that the SG is potentially filled solid with water. It would be correct that a transition to FR-H.3 would be required if the PAM grade instruments were in excess of the setpoint; however, in this case they are not and as such the crew will remain in FR-H.2.

D. Incorrect: While it is true that the ICS data suggests that the SG is potentially filled solid, it is not true (as previously discussed) that the data can be used to make a transition to FR-H.3.

Question Number: 85 Tier: 1 Group: 2 K/A: WE13 Steam Generator Over-pressure G2.4.3 Ability to identify post-accident instrumentation Importance Rating: 3.7 3.9 10 CFR Part 55: (CFR: 41.6 / 45.4) 10CFR55.43.b: 10 CFR 55.43(b)(5)

K/A K/A is matched because the applicant is required to correctly Match: implement the steps of FR-H.2 given displays of both PAM instrumentation and regular instrumentation.

Technical

Reference:

FR-H.2, Steam Generator Overpressure FR-H.3, Steam Generator High Level TI-12.04, Users Guide for Abnormal and Emergency Operating Instructions Proposed references None to be provided:

Learning Objective: 3-OT-FRH0001

3. Explain the purpose for and the basis of each step in FR-H procedures
6. Given a set of plant conditions use the FR-H procedures to correctly identify and required procedure transition Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question for the 2015-301 NRC SRO Exam Comments: See the marked up Clarification Guidance for SRO-only Questions.

STEP DESCRIPTION TABLE FOR FR-H.2 Step 3__

STEP: Check Affected SG(s) Narrow Range Level - LESS THAN (M.08)%

[(M.09)% FOR ADVERSE CONTAINMENT]

93% for WBN PURPOSE: To determine if overfilling the affected SG is a potential cause of the overpressurization BASIS:

The operator should check the affected SG level to ensure that it is not above (M.08)% [(M.09)% for adverse containment]. If the level is greater than (M.08)% [(M.09)% for adverse containment], the SG water level may be above the narrow range or the SG may even be filled solid with water. For this case, the operator is transferred to FR-H.3, RESPONSE TO STEAM GENERATOR HIGH LEVEL, to address the high water level condition. At 94%, the S/G is potentially filled ACTIONS: solid with water.

o Determine if the affected SG narrow range level is less than (M.08)%

[(M.09)% for adverse containment]

o Transfer to FR-H.3, RESPONSE TO STEAM GENERATOR HIGH LEVEL, Step 1 INSTRUMENTATION:

SG narrow range level CONTROL/EQUIPMENT:

N/A KNOWLEDGE:

N/A PLANT-SPECIFIC INFORMATION:

o (M.08) SG level at the upper tap, including allowances for normal channel accuracy.

o (M.09) SG level at the upper tap, including allowances for normal channel accuracy, post-accident transmitter errors, and reference leg process errors.

FR-H.2 Background 11 HP-Rev. 2, 4/30/2005 HFRH2BG.doc

WBN User's Guide for 0-TI-12.04 Unit 1 & 2 Abnormal and Emergency Rev. 0000 Operating Instructions Page 12 of 57 2.2.1 Cautions and Notes (continued)

C. CAUTIONS and NOTES are introduced by their designator in bold face type.

1. The designator is followed by the text extending across the entire page with note text appearing in standard type and caution text appearing in bold face type.
2. If multiple cautions or notes are applicable to a step then each caution or note included after the initial designator is distinguished by a preceding bullet.

D. In general, CAUTIONS and NOTES apply to the step which they precede.

E. CAUTIONS and NOTES which precede the first operator action step may also apply throughout the instruction.

F. When CAUTIONS or NOTES are communicated (Read by procedure reader) they are to be communicated through directive communication addressed to individual(s) and verified via 3-way communication.

2.2.2 Use of Instrumentation A. Post Accident Monitoring (PAM) instrumentation is provided on the main control board as determined by design requirements.

1. The control room complies with Reg. Guide 1.97 requirements by providing Because PAM S/G the operator with the required PAM instruments.

NR is provided on the MCR board 1- 2. The control board PAM instruments are uniquely labeled to identify them M-4, it is required as PAM instrumentation.

to be used.

a. Most PAM instruments have black background instrumentation labels.
b. Other PAM instruments are identified by a small box located on the instrument label with the designator C1 or C2 inside the box.
3. During performance of the EOI set, the operator is required to utilize PAM instruments when they are provided on the control board.
4. The operator should compare redundant instruments when they are provided.
5. Some parameters evaluated during performance of the EOIs do NOT have PAM grade instrumentation provided on the control board.
a. The operator should monitor these parameters with available instrumentation.

WBN Steam Generator Overpressure FR-H.2 Unit 1 Rev. 0006 Step Action/Expected Response Response Not Obtained 3.0 OPERATOR ACTIONS

1. IDENTIFY affected S/G(s):
a. Any S/G pressure greater a. IF press in all S/Gs less than than or equal to 1220 psig. 1220 psig, THEN RETURN TO Instruction in effect.
2. ENSURE MFW isolated to Manually CLOSE valves, AND affected S/Gs:

STOP pumps, as necessary.

  • S/G MFW isolation and bypass isolation valves CLOSED.
  • S/G MFW reg and bypass reg IF valves can NOT be closed, THEN valves CLOSED.

CLOSE #1 heater outlet valves.

  • MFP A and B TRIPPED.

This is the

  • Standby MFP STOPPED. transition being
  • Cond demin pumps TRIPPED. tested.
  • Cond booster pumps TRIPPED.
3. CHECK affected S/Gs NR level ** GO TO FR-H.3, STEAM less than 93% [85% ADV]. GENERATOR HIGH LEVEL.

The adverse

4. DEPRESSURIZE affected S/Gs:

setpoint is not in

  • OPEN S/G PORVs, effect for this question.

OR

  • OPEN S/G MSIV bypass valves, Because the PAM instruments exist on 1-M-4 for SG NR level, they must be OR used. The transition to FR-H.3 cannot be made using the ICS value. It must
  • OPEN S/G steam supply be made using the PAM instruments' to TD AFW pump, readings. This is seen in TI-12.04.

OR

  • OPEN S/G blowdown valves.

Page 3 of 6

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]

Some examples of SRO exam items for this topic include:

  • Administrative processes for the installation of temporary instrumentation.
  • Processes for changing the plant or plant procedures.

Section IV provides an example of a satisfactory SRO-only question related to this topic.

D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

[10 CFR 55.43(b)(4)]

Some examples of SRO exam items for this topic include:

  • Process for gaseous/liquid release approvals, i.e., release permits.
  • Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.

Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.

SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.

E. Assessment of facility conditions and selection of appropriate procedures The applicant must during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

assess the condition of the SG This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or by using both PAM section of a procedure to mitigate, recover, or with which to proceed. One and regular area of SRO level knowledge (with respect to selecting a procedure) is instrumentation knowledge of the content of the procedure versus knowledge of the and then procedures overall mitigative strategy or purpose. Detailed implement the knowledge of the correct procedure. The applicants knowledge can be evaluated at the level of 10 CFR transitional step is 55.43(b)(5) by ensuring that the additional knowledge of the procedures required (i.e.

content is required to correctly answer the written test item, for example:

knowledge of the setpoint).

Page 6 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

  • Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
  • Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
  • Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.

SRO-only knowledge should not be claimed for questions that can be answered solely using systems knowledge; e.g.:

  • how the system works. The procedures
  • system flow path. FR-H.2 and FR-H.3
  • component locations, etc. are both yellow path functional SRO-only knowledge should not be claimed for questions that can restorations.

be answered solely using fundamental knowledge of:

  • the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
  • any AOP entry condition.
  • plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
  • immediate operator actions of a procedure.

Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.

Page 7 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, Yes RO question flowpath, logic, component location?

No Can the question be answered solely by knowing immediate operator actions? Yes RO question No Can the question be answered solely by knowing entry conditions for AOPs or plant parameters Yes RO question that require direct entry to major EOPs?

No Can the question be answered solely by knowing the purpose, overall sequence of events, or Yes RO question overall mitigative strategy of a procedure?

No Does the question require one or more of the following?

  • Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
  • Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps Yes SRO-only
  • Knowledge of diagnostic steps and decision points in the question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
  • Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

86.

Given the following timeline:

00:00:00 - Unit 1 is at 100%.

00:0 1:00 - The primary +15VDC power supply trips in train B SSPS.

- The secondary +15VDC power supply picked up the SSPS logic circuit card load.

- SSPS-B GEN WARNING (115-A) is LIT.

06:00:00 - A troubleshooting/repair work order is authorized and placed in WORKING status.

08:00:00 - The SSPS engineer wishes to block the General Warning signal from the failed +15VDC power supply to facilitate troubleshooting.

- He proposes performing the following under the work order authorized at 06:00:00:

o Lift the output wires for the failed +15VDC power supply.

o Install a jumper to supply the failed input to the general warning circuit from the functional +15VDC power supply.

o The jumper is expected to be installed for two weeks.

Which ONE of the following describes the operability of the B Train SSPS AND whether or not a Technical Evaluation (TE) review is required?

In accordance with T/S LCO 3.3.1, RTS Instrumentation B Train SSPS is ____(1)____ at 00:0 1 :0 1.

Performance of the recommendations made at 08:00:00 ____(2)____ require a TE review In accordance with NPG-SPP-09.5, Temporary Modifications.

REFERENCE PROVIDED A. (1) operable (2) WILL B. (1) operable (2) WILL NOT C. (1) inoperable (2) WILL D. (1) inoperable (2) WILL NOT

CORRECT ANSWER: A DISTRACTOR ANALYSIS:

A. Correct: The Westinghouse manual for the SSPS system, WBN-VTD-W120-2454, indicates that two 15VDC power supplies exist within the logic bay of each SSPS train. These supplies feed redundantly to the 15VDC buses; they do this through the action of an auctioneering circuit. This design promotes the continuity of power of the generating unit should a solitary power supply be lost. Because of this design, there is no impact to the operability of SSPS given the loss of a single low voltage power supply. Therefore, it is correct that the B SSPS train was operable immediately following the loss of one of the two

+15VDC power supplies.

It is correct that a TE be required to implement the desires of the system engineer. This is seen in section 3.7 E which states: A Technical Evaluation (TE) review is required for all WO-TMs.

B. Incorrect: While it is correct that the B SSPS train remains operable following the loss of a solitary +15VDC power supply, it is not correct that a Technical Evaluation not be required. It is plausible (for many reasons) why this evaluation would not be required. Firstly, if the applicant may believe that the exclusion 2.2 I.2 of the aforementioned SPP applied. This exclusion states that Connections to permanently installed test jacks to take a reading are excluded from the restrictions of the procedure. The proposed connections for the jumper to be installed are test jacks and are normally used to measure the output of the two 15VDC power supplies. However, his proposal is not to conduct measurements but to cross connect power supplies; therefore, the exclusion is not applicable. Next, (as seen in 2.2 K),

Temporary changes that are continuously attended are excluded.

However, the stem of the question details that the jumper will not be attended and as such is not allowed the liberty of the described exclusion.

C. Incorrect: While it is correct that the jumper does require that a technical evaluation review be conducted, it is not correct that the train of SSPS is rendered inoperable following the failure of a +15VDC power supply.

D. Incorrect: As discussed it is incorrect and yet plausible that the SSPS train is rendered inoperable upon the loss of one of the two +15VDC power supplies. Also, it is incorrect and yet plausible that a Technical Evaluation would not be required.

Question Number: 86 Tier: 2 Group: 1 K/A: 012 Reactor Protection System 2.2 Equipment Control 2.2.36 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

Importance Rating: 3.1 4.2 10 CFR Part 55: (CFR: 41.10 / 43.2 / 45.13) 10CFR55.43.b: 10 CFR 55.43(b)(2) and 10 CFR 55.43(b)(3)

K/A Match: K/A is matched because the applicant is required to analyze the effect that a degraded 15VDC power supply has on a SSPS train using both systems knowledge and the contents of the T/S bases.

Technical

Reference:

PER Vault Summary for PER 3516 Bases for T/S LCO 3.3.1 NPG-SPP-09.5, Temporary Modifications Westinghouse SSPS Technical Manual, WBN-VTD-W120-2454 50.59 screen for WO 01-008855-000 Proposed references to None be provided:

Learning Objective:

Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question for the 2015-301 NRC SRO Exam Comments: See the marked up Clarification Guidance for SRO-only Questions.

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]:

A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)]

Some examples of SRO exam items for this topic include:

  • Reporting requirements when the maximum licensed thermal power output is exceeded.
  • The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
  • National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
  • Processes for TS and FSAR changes.

Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic.

B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

The question Some examples of SRO exam items for this topic include:

requires the applicant to

  • Application of Required Actions (Section 3) and Surveillance determine whether Requirements (SR) (Section 4) in accordance with rules of application or not a train of requirements (Section 1).

SSPS remains

  • Application of generic Limiting Condition for Operation (LCO) operable following requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).

a loss of a power

  • Knowledge g of TS bases that are required to analyze TS required actions supply. and terminology.

SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of  1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.

SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. ROs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Page 3 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)

RO knowledge Above this line Page 4 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)

(Tech Specs)

Can question be answered solely by knowing  1 Yes hour TS/TRM Action? RO question No Can question be answered solely by knowing the Yes LCO/TRM information listed above-the-line? RO question No Can question be answered solely by knowing the Yes TS Safety Limits? RO question No Does the question involve one or more of the following for TS, TRM, or ODCM?

  • Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
  • Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)) Yes SRO-only
  • Knowledge g of TS bases that is required to analyze TS question required actions and terminology No Therefore, the determination of Question might not be linked to operability of the 10 CFR 55.43(b)(2) for SRO-only train of SSPS is SRO only.

Page 5 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

C. Facility y licensee pprocedures required q to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]

The use of NPG-Some examples of SRO exam items for this topic include: SPP-09.5 meets the intent of this

  • Administrative processes for the installation of temporary instrumentation.
  • Processes for changing the plant or plant procedures.

Section IV provides an example of a satisfactory SRO-only question related to this topic.

D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

[10 CFR 55.43(b)(4)]

Some examples of SRO exam items for this topic include:

  • Process for gaseous/liquid release approvals, i.e., release permits.
  • Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.

Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.

SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.

E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.

The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:

Page 6 of 16

RTS Instrumentation B 3.3.1 Bases APPLICABLE 18. Reactor Trip Breaker Undervoltage and Shunt Trip SAFETY ANALYSES, Mechanisms (continued)

LCO, and APPLICABILITY service. The trip mechanisms are not required to be OPERABLE for trip breakers that are open, racked out, incapable of supplying power to the CRD System, or declared inoperable under Function 17 above.

OPERABILITY of both trip mechanisms on each breaker ensures that no single trip mechanism failure will prevent opening any breaker on a valid signal.

These trip Functions must be OPERABLE in MODE 1 or 2 when the reactor is critical. In MODE 3, 4, or 5, these RTS trip Functions must be Note that the SSPS OPERABLE when the RTBs and associated bypass breakers are closed, remains completely and the CRD System is capable of rod withdrawal.

capable of

19. Automatic Trip Logic performing this function following The LCO requirement for the RTBs (Functions 17 and 18)and Automatic the failure of the Trip Logic (Function 19) ensures that means are provided to interrupt the single power power to allow the rods to fall into the reactor core. Each RTB is equipped with an undervoltage coil and a shunt trip coil to trip the breaker supply.

open when needed. Each RTB is equipped with a bypass breaker to allow testing of the trip breaker while the unit is at power. The reactor trip signals generated by the RTS Automatic Trip Logic cause the RTBs and associated bypass breakers to open and shut down the reactor.

The LCO requires two trains of RTS Automatic Trip Logic to be OPERABLE. Having two OPERABLE channels ensures that random failure of a single logic channel will not prevent reactor trip.

These trip Functions must be OPERABLE in MODE 1 or 2 when the reactor is critical. In MODE 3, 4, or 5, these RTS trip Functions must be OPERABLE when the RTBs and associated bypass breakers are closed, and the CRD System is capable of rod withdrawal.

The RTS instrumentation satisfies Criterion 3 of the NRC Policy Statement.

(continued)

Watts Bar-Unit 1 B 3.3-38

The power supplies are redundant.

PER Vault Summary Report for PER: 3516 PER Number: 3516 Status: ARCHIVE Status Date:

Status By ID:

Site / Org: WBN PER Level: CAP Due Date:

Unit: Date of Occurence: Long Lead CA Date:

Analysis Type:

PER Summary:

Problem Evaluation Report - User Information Originator ID: On Behalf of ID:

Originator Name: On Behalf of Name:

Originator Phone: On Behalf of Phone:

Originator Email: On Behalf of Email:

Problem Evaluation Report - Problem Details Initiating Department: Asset:

Site / Org: WBN Asset Site: The source of the SBU: Location:

BU: Additional Location Details:

scenario for the Unit: Plant System: question.

Corrected Immediately: NO Reported Date:

Date Of Occurence:

PER Summary:

PER Details: ONE 15VDC POWER SUPPLY FOR SSPS TR B,, LOCATED IN 1-R-50,, HAS GONE BAD. THIS BROUGHT IN A GENERAL WARNING FOR B TR SSPS. B TR SSPS IS STILL OPERABLE DUE TO THE AUCTIONEERING CIRCUIT FOR THE 15VDC PWR SUPPLIES.

Immediate Actions Taken:

Actions Taken Details:

Previous Site ID:

Site Change Flag: NO Problem Evaluation Report - Customer Impact Outage

Reference:

As Found Condition: Customer

Reference:

Problem Evaluation Report - Regulatory Impact Potential Environmental Issue: NO Potential Operability Issue: Potential Reactivity Issue: NO Potential Reportability Issue: NO Potential Safety Issue: NO Management Screening - Review Results / Approval PER Level: Tier: PER Category: Good Catch: NO Potential Margin Issue: NO Justification:

Justification Details:

Responsible Org: Analysis Type: CAP Due Date:

Comments:

Comment Details:

Page 1 of 25 TVA RESTRICTED INFORMATION Wednesday, December 10, 2014

PER Vault Summary Report for PER: 3516 Regulatory Reviews - Environmental Review Event Number: Event

Description:

Environmental Issue?: Environmental Type: Event Repeat?: NO Potential Environmental Issue?: NO Source Code: Environmental Area: Notice of Violation?: NO Non-Conformance Code: Findings Code: EMS Process:

Justification for Environ. Disposition:

Justification Details:

Reviewer ID:

Review Date:

Regulatory Reviews - Safety Review Safety Event Number: Potential Safety?: NO Violation Notice Required?: NO Source Code:

Safety Issue?: Findings Code:

Justification for Safety Disposition:

Justification Details:

Reviewer ID:

Review Date:

Regulatory Reviews - Operations Review Issue?

Potential Operability Issue?: NO Operability Issue?: NO The crew screened Operability Actions:

the issue Operability Actions Details:

Potential Reportability Issue?: NO (correctly) as not Reportability Issue?: NO an impact to the Engineering Evaluation Needed?: operability of Required Date:

SSPS.

Ops Notified Other Sites?: NO Operability Reviewer:

Operability Review Date:

Reportability Reviewer:

Reportability Review Date:

Regulatory Reviews - Engineering Evaluation Engineering Evaluation Needed?:

Required Date:

Evaluation Summary:

Evaluation Details:

Outside CLB?: NO CLB Affected:

CLB Affected Details:

Specified (Safety) Function Maintained?: NO Immediate/Compensatory Measures?: NO Page 3 of 25 TVA RESTRICTED INFORMATION Wednesday, December 10, 2014

Not an impact to operability.

NPG Standard Temporary Modifications NPG-SPP-09.5 Programs and Temporary Configuration Changes Rev. 0009 Processes Page 18 of 77 3.6 Procedurally Controlled Temporary Modifications (PCTM) (continued)

C. The procedure will include the following administrative information:

This is the category 1. The section of the procedure that implements the PCTM will be clearly identified of Tmod described as a PCTM.

by the question.

2. The section will include a note that any changes will require a 50.59 / 72.48 evaluation, a Technical Evaluation, and the Design Control review. (except for non-intent changes - minor or editorial)
3. If the PCTM is installed greater than one shift, the modification will be tagged and entered in the Temporary Modification Log.

3.7 Temporary Modifications in Support of Maintenance (WO-TM)

A. Temporary modification in support of maintenance are implemented under 10 CFR 50.65 (Maintenance Rule) rather than 10 CFR 50.59 (Changes, Tests, and Experiments). The modification may remain installed for 90 days at power under 50.65. Beyond 90 days at power, a 50.59 / 72.48 review is required.

B. The modification must meet the following criteria to be processed as a WO-TM:

1. The modification must be in direct support of maintenance (for example, Therefore, a TE is necessary to establish work conditions or provide equipment necessary to required for the perform work - see Attachment 19 for further clarification)

Tmod proposed by the question. 2. The modification must be controlled by an active WO.

3. The modification must not impact the decision-making capability of the plant Operators and must not require changes to operating procedures or Operator training.

C. A Technical Evaluation (TE) review is required for all WO-TMs.

1. Engineering will perform a Technical Evaluation in accordance with the Technical Evaluation Form (Attachment 3) and the step-by-step instructions in Attachment 21.
2. Engineering will provide the Technical Evaluation to Planning who will incorporate any special instructions or requirements into the WO package.
3. The Technical Evaluation will be included in the WO package.

D. A 50.59 / 72.48 review is required under the following conditions:

1. If plant personnel expect the modification will be installed less than 90 days, then a 50.59 / 72.48 review is not required.
2. If plant personnel know beforehand that the modification will be in place for more than 90 days, then a 50.59 / 72.48 review is required.

87.

Given the following conditions:

- Unit 1 is in MODE 3.

Subsequently:

- A Loss of 120v AC Vital Instrument Power Board 1-III occurs.

- 1-AOI-25.03, Loss of 120V AC Vital Instrument Power Boards 1-III or 2-III, directs use of 1-SOI-235.03, 120V AC Vital Power System 1-III, to restore the board Which ONE of the following describes the status of the system AND which procedure will allow exiting the T/S LCO?

During the loss of the vital board, ____(1)____ Unit 1 SSPS Train A ESF relays COULD be energized.

Implementing section ____(2)____ of 1-SOI-235.03 would re-energize the board and allows BOTH T/S LCOs 3.8.7, Inverters and 3.8.9, Distribution Systems - Operating to be exited.

A. (1) ONLY the master (2) 8.1, Transferring 480V AC Vital Transfer Switch III to Alternate 480V Power Supply B. (1) BOTH the master and the slave (2) 8.1, Transferring 480V AC Vital Transfer Switch III to Alternate 480V Power Supply C. (1) ONLY the master (2) 8.3, Transfer 120V AC Vital Instrument Power Board 1-III to Spare 120V AC Vital Inverter 0-III D. (1) BOTH the master and the slave (2) 8.3, Transfer 120V AC Vital Instrument Power Board 1-III to Spare 120V AC Vital Inverter 0-III

CORRECT ANSWER: D DISTRACTOR ANALYSIS:

A. Incorrect: As seen on the simplified Westinghouse drawing, the SSPS power supply distribution involves the provision of 120V AC Vital Instrumentation Buses I and III to the A train SSPS and Buses II and IV to the B train SSPS. One may observe on this simplified drawing that Buses I and III each power a 48V and 15V power supply pair.

The 48V power supplies are auctioneered together to provide a 48V bus within the SSPS rack. One of the users of this buss power is the operating coils for the SSPS master relays. Therefore, the loss of one of the two 120V AC Vital Instrument Buses which is provided to a train of SSPS will not cause a loss of power to the SSPS master relays because the remaining Vital Instrument Bus will continue to supply the redundant feed of 48V power. Another feed from Bus I to the A train SSPS is the provision of 120V AC power to the slave relay coils. One must note that there is no redundant supply of AC power to these slave relays. Therefore, if Bus I is lost, the A train SSPS slave relays will be depowered and thus unable to actuate. Germane to this question is that the loss of Bus III 120V AC will cause no impact to the slave relays. Therefore, upon the loss of Bus III 120V AC, both the master and slave relays could be energized.

It is plausible to believe that the slave relays could not be energized provided that one believed that Bus III powered the slave relay coils.

As seen in the print excerpt taken from 1-45W700-1, the 120V AC Vital Instrument buses are provided with power from an inverter unit. This unit can provide power from three basic sources: 1. power can be provided from an inverter which is supplied DC from the battery board,

2. power can be provided from an inverter which is supplied from an AC feed which is rectified and 3. power can be provided from a transformed and regulated AC feed which bypasses the inverter completely.

From the perspective of operability, one may see that T/S LCO 3.8.9 allows any one of the three aforementioned sources of power to supply a 120V AC bus. One may further see that T/S LCO 3.8.7 is different in that it does not allow the transformed an regulated AC feed to be utilized. Therefore, the use of section 8.1 (which would place the transformed an regulated AC feed in service) would NOT allow T/S LCO 3.8.7 to be satisfied. The plausibility to this distractor is lent by the fact that T/S LCO 3.8.9 allows the use of the bypass feed while T/S LCO 3.8.7 does not.

Note that both T/S LCO 3.8.7 and 3.8.9 are applicable in Modes 1-4.

B. Incorrect: While it is correct that both the master and slave relays would be able

to be energized, it is not correct and yet plausible that the use of the bypass AC feed would allow T/S LCO 3.8.7 to be MET.

C. Incorrect: It is incorrect and yet plausible that only the master relays could be energized. It is correct (as discussed), that the transfer of the AC Vital Board to the spare inverter would allow both T/S LCO 3.8.9 and T/S LCO 3.8.7 to be exited.

D. Correct: As noted, both of the parts of this answer are correct.

Question Number: 87 Tier: 2 Group: 1 K/A: 013 Engineered Safety Features Actuation System A2 Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS; and (b) based Ability on those predictions, use procedures to Correct: control, or mitigate the consequences of those malfunctions or operations; Importance Rating: A2.04 Loss of instrument bus 10 CFR Part 55: (CFR: 41.5 / 43.5 / 45.3 / 45.13) 10CFR55.43.b: 10 CFR 55.43(b)(2)

K/A Match: The K/A is matched because the question requires predicting the impact of the loss if an instrument bus on the ESFAS. The question then requires selecting the correct section of an SOI to both restore power and maintain T/S LCO operability.

Technical

Reference:

1-45W700-1 Simplified Westinghouse graphic showing the Power Distribution to SSPS 1-SOI-235.03, 120V AC Vital Power System 1 III T/S Basis for LCO 3.8.7, Inverters - Operating T/S Basis for LCO 3.8.9, Distribution Systems - Operating Proposed references to None be provided:

Learning Objective: 3-OT-SYS235A

4. EXPLAIN the physical connections and/or cause-effect relationships between the 120 Volt AC System and the following:
a. Solid State Protection System (SSPS)
10. DESCRIBE the following aspects of TS and TRs
b. The Limiting Conditions for Operation, Applicability, and Bases.

Cognitive Level:

Higher X Lower Question Source:

New Modified Bank Bank X Question History: Bank Question 013 A2.04 87 which was used on the 06/2011 WBN NRC exam.

Comments: The question meets the general SRO only criteria of Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

SSPS Power Supply Distribution Bus I 120V AC Vital Instrumentation Bus II Bus III Bus IV From Train A From Train B Safeguards Safeguards Train A Test Panel Train B Test Panel I I 48v 15v To Slave To Slave Relays Relays II II 48v 15v III III 48v 15v IV IV 48v 15v

From an inverter Distribution Systems - Operating (normal or spare) B 3.8.9 from either 1. the inverter whenBASES powered by the LCO Maintaining the Train A and Train B AC, four channels of vital DC, and battery, 2. the (continued) four channels of AC vital bus electrical power distribution subsystems inverter when OPERABLE ensures that the redundancy incorporated into the design of ESF is powered by the ac to not defeated. Therefore, a single failure within any system or within the electrical dc converter, or 3. power distribution subsystems will not prevent safe shutdown of the reactor.

the isolimiter OPERABLE AC electrical power distribution subsystems require the associated (regulated buses, load centers, motor control centers, and distribution panels to be transformer bypass energized to their proper voltages. OPERABLE DC electrical power distribution source) subsystems require the associated buses to be energized to their proper voltage from either the associated battery or charger. OPERABLE vital bus electrical power distribution subsystems require the associated buses to be energized to their proper voltage from the associated unit or spare inverter via inverted DC THIS IS DIFFERENT voltage, unit or spare inverter using internal AC source, or the regulated THAN T/S LCO 3.8.7 transformer bypass source.

In addition, tie breakers between redundant safety related AC, vital DC, and AC vital bus power distribution subsystems, if they exist, must be open. This prevents any electrical malfunction in any power distribution subsystem from propagating to the redundant subsystem, that could cause the failure of a redundant subsystem and a loss of essential safety function(s). If any tie breakers are closed, the affected redundant electrical power distribution subsystems are considered inoperable. This applies to the onsite, safety related redundant electrical power distribution subsystems. It does not, however, preclude redundant 6.9 kV shutdown boards from being powered from the same offsite circuit.

APPLICABILITY The electrical power distribution subsystems are required to be OPERABLE in MODES 1, 2, 3, and 4 to ensure that:

a. Acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result of AOOs or abnormal transients; and
b. Adequate core cooling is provided, and containment OPERABILITY and other vital functions are maintained in the event of a postulated DBA.

Electrical power distribution subsystem requirements for MODES 5 and 6 are covered in the Bases for LCO 3.8.10, "Distribution Systems - Shutdown."

(continued)

Watts Bar-Unit 1 B 3.8-91 Revision 67, 75, 76, 77, 78

Inverters - Operating B 3.8.7 BASES APPLICABLE Inverters are a part of the distribution systems and, as such, satisfy Criterion 3 SAFETY ANALYSIS of the NRC Policy Statement.

(continued)

LCO The inverters ensure the availability of AC electrical power for the systems instrumentation required to shut down the reactor and maintain it in a safe condition after an anticipated operational occurrence (A00) or a postulated DBA.

Maintaining the required inverters OPERABLE ensures that the redundancy The inverter is incorporated into the design of the RPS and ESFAS instrumentation and controls operable when the is maintained. The twelve inverters (one Unit 1, one Unit 2 and one spare per inverter is channel) ensure an uninterruptible supply of AC electrical power to the AC vital buses even if the 6.9 kV shutdown boards are de-energized.

supplying the vital AC board (not the OPERABLE inverters require the associated AC vital bus to be powered by an regulated bypass inverter with output voltage and frequency within tolerances and power input to supply). the inverter from a 125 VDC vital battery. Alternatively, power supply may be from an internal AC source via rectifier as long as the vital battery is available as the uninterruptible power supply. The unit inverters have an associated bypass supply provided by a regulated transformer that is automatically connected to the associated AC vital bus in the event of inverter failure or overload. The bypass supply is not battery-backed and thus does not meet requirements for inverter operability. The spare inverters do not have an associated bypass supply.

Additionally, the inverter channel must not be connected to the cross train 480 V power supply.

APPLICABILITY The inverters are required to be OPERABLE in MODES 1, 2, 3, and 4 to ensure that:

a. Acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result of AOOs or abnormal transients; and
b. Adequate core cooling is provided, and containment OPERABILITY and other vital functions are maintained in the event of a postulated DBA.

Inverter requirements for MODES 5 and 6 are covered in the Bases for LCO 3.8.8, "Inverters - Shutdown."

(continued)

Watts Bar-Unit 1 B 3.8-82 Revision 58, 67, 75, 76, 77, 78, 97 Amendment 45, 76

WBN 120V AC Vital Power System 1-III 1-SOI-235.03 Unit 1 Rev. 0000 Page 25 of 35 Date________ Initials This section renders T/S LCO 3.8.7 NOT MET 8.0 INFREQUENT OPERATIONS 8.1 Transferring 480V AC Vital Transfer Switch III to Alternate 480V Power Supply CAUTIONS

1) Consideration should be given to a possible loss of Channel 3 SSPS and ESF should 120V AC Vital Power Board 1-III and 2-III lose potential.
2) EMERGENCY feeder from 480V SHUTDOWN BOARD 2B1-B on 0-XSW-236-3, 480V AC VITAL TRANSFER SWITCH III, is NOT accounted for in D/G loading calculations and shall NOT be used without engineering evaluation (see note 9, drawing 1-15E500-2).

[1] OBTAIN current approved engineering evaluation for this performance and attach a copy to this Data Package. ________

SRO

[2] CHECK 1-EI-235-3/V2, BATTERY INPUT on Inverter 1-III OR 0-EI-235-3/V2, BATTERY INPUT on Inverter 0-III, if in service, to be 133-140 VDC. ________

[3] IF INSTRUMENT POWER A RACK TRANSFER SWITCH, in NORMAL FEEDER, THEN PLACE INSTRUMENT POWER A RACK TRANSFER SWITCH, in ALTERNATE FEEDER. [2-M-7] ________

CV

[4] IF INSTRUMENT POWER B RACK TRANSFER SWITCH, in ALTERNATE FEEDER, THEN PLACE INSTRUMENT POWER B RACK TRANSFER SWITCH, in NORMAL FEEDER. [2-M-7] ________

CV

WBN 120V AC Vital Power System 1-III 1-SOI-235.03 Unit 1 Rev. 0000 Page 26 of 35 Date________ Initials 8.1 Transferring 480V AC Vital Transfer Switch III to Alternate 480V Power Supply (continued)

[5] EVALUATE possible effects on all feeds from 120V Instrument Power Distribution Panel 2-A, including feeds from Panel 2-M-7, due to momentary loss of potential, to include the following systems:

Access control system -Nuclear Security- Momentary loss if on alternate supply (notify Nuclear Security). ________

[6] CHECK status windows for other ESF or SSPS channels LIT which could cause a Reactor Trip or Safety Injection should channel 1 be lost on the power supply transfer. ________

[7] CLOSE 0-BKR-236-3A, ALT FDR FOR VITAL BATT CHGR III (0-CHGR-236-3), on 480V SHUTDOWN BOARD 2B1-B

[C/10A]. ________

CV

[8] CHECK 0-XSW-236-3, 480V AC VITAL TRANSFER SWITCH III, EMERGENCY (supply feeder) red light ON. ________

[9] PLACE 0-XSW-236-3, 480V AC VITAL TRANSFER SWITCH III, NORMAL supply to OFF. ________

CV

[10] PLACE 0-XSW-236-3, 480V AC VITAL TRANSFER SWITCH III, EMERGENCY supply to ON. ________

CV

WBN 120V AC Vital Power System 1-III 1-SOI-235.03 Unit 1 Rev. 0000 Page 27 of 35 Date________ Initials 8.1 Transferring 480V AC Vital Transfer Switch III to Alternate 480V Power Supply (continued)

[11] CHECK the following equipment ENERGIZED:

[11.1] 120V AC VITAL INVERTER 1-III or 0-III, if in service ________

[11.2] 120V AC VITAL INSTR POWER BOARD 1-III. ________

[11.3] 120V AC VITAL INSTR POWER BOARD 2-III. ________

[11.4] 125-V VITAL BATTERY CHARGER III. ________

[12] PLACE INSTRUMENT POWER A RACK TRANSFER SWITCH, in NORMAL FEEDER. [2-M-7] ________

CV

[13] CHECK 0-XSW-236-3, 480 V AC VITAL TRANSFER SWITCH III, NORMAL supply to OFF. ________

IV End of Section

WBN 120V AC Vital Power System 1-III 1-SOI-235.03 Unit 1 Rev. 0000 Page 28 of 35 Date________ Initials 8.2 Transferring 480V AC Vital Transfer Switch III to Normal 480V Power Supply CAUTION Consideration should be given to a possible loss of Channel 3 SSPS and ESF should 120V AC Vital Power Board 1-III and 2-III lose potential.

[1] OBTAIN SRO approval prior to performing this Section. ________

SRO

[2] CHECK 1-EI-235-3/V2, BATTERY INPUT on Inverter 1-III OR 0-EI-235-3/V2, BATTERY INPUT on Inverter 0-III, if in service, to be 133-140 VDC. ________

[3] IF INSTRUMENT POWER A RACK TRANSFER SWITCH, in NORMAL FEEDER, THEN PLACE INSTRUMENT POWER A RACK TRANSFER SWITCH, in ALTERNATE FEEDER. [2-M-7] ________

CV

[4] IF INSTRUMENT POWER B RACK TRANSFER SWITCH, in ALTERNATE FEEDER, THEN PLACE INSTRUMENT POWER B RACK TRANSFER SWITCH, in NORMAL FEEDER. [2-M-7] ________

CV

[5] EVALUATE possible effects on all feeds from 120V Instrument Power Distribution Panel 2-A, including feeds from Panel 2-M-7, due to temporary loss of potential, to include the following systems:

Access control system -Nuclear Security- Momentary loss if on alternate supply (notify Nuclear Security). ________

[6] CHECK status windows for other ESF or SSPS channels LIT which could cause a Reactor Trip or Safety Injection should channel III be lost on the power supply transfer. ________

WBN 120V AC Vital Power System 1-III 1-SOI-235.03 Unit 1 Rev. 0000 Page 29 of 35 Date________ Initials 8.2 Transferring 480V AC Vital Transfer Switch III to Normal 480V Power Supply (continued)

[7] CHECK 0-XSW-236-3, 480V AC VITAL TRANSFER SWITCH III, NORMAL (supply feeder) red light ON. ________

[8] PLACE 0-XSW-236-3, 480 V AC VITAL TRANSFER SWITCH III, EMERGENCY supply to OFF. ________

CV

[9] PLACE 0-XSW-236-3, 480 V AC VITAL TRANSFER SWITCH III, NORMAL supply to ON. ________

CV

[10] CHECK the following equipment ENERGIZED:

[10.1] 120V AC VITAL INVERTER 1-III or 0-III, if in service. ________

[10.2] 120V AC VITAL INSTR POWER BOARD 1-III. ________

[10.3] 120V AC VITAL INSTR POWER BOARD 2-III. ________

[10.4] 125-V VITAL BATTERY CHARGER III. ________

[11] PLACE INSTRUMENT POWER A RACK TRANSFER SWITCH, in NORMAL FEEDER. [2-M-7] ________

CV

[12] OPEN 0-BKR-236-3A, ALT FDR FOR VITAL BATT CHGR III (0-CHGR-236-3), on 480V SHUTDOWN BOARD 2B1-B

[C/10A]. ________

CV

[13] CHECK 0-XSW-236-3, 480 V AC VITAL TRANSFER SWITCH III, EMERGENCY supply to OFF. ________

IV End of Section

WBN 120V AC Vital Power System 1-III 1-SOI-235.03 Unit 1 Rev. 0000 Page 30 of 35 Date________ This section allows both Initials T/S LCO 3.8.7 and T/S 8.3 Transfer 120V AC Vital Instrument Power Board 1-III to Spare LCO 3.8.9 to be MET.

120V AC Vital Inverter 0-III NOTE This procedure section transfers the 120V AC power supply to 120 V AC Vital Instrument Power Board 1-III from 120V AC Vital Inverter 1-III to Spare 120V AC Vital Inverter 0-III.

This section is not used to energize a dead board. Refer to Sections 5.2 and 5.3 to energize dead 120V AC Vital Instrument Power Board 1-III using Spare 120 V AC Vital Inverter 0-III.

[1] ENSURE Spare Inverter 0-III has been placed in service per Section 5.2, Startup of 120V AC Vital Inverter 0-III. ________

CV

[2] ENSURE 120V AC VITAL INVERTER SUPPLY AVAILABLE amber light on 120V AC VITAL INSTR POWER BD 1-III LIT. ________

CV

[3] ENSURE 120V AC ALTERNATE SUPPLY AVAILABLE amber light on 120V AC VITAL INSTR POWER BD 1-III LIT. ________

CV

[4] ENSURE 120V AC VITAL INVERTER & ALT SUPPLY IN SYNC blue light on 120V AC VITAL INSTR POWER BD 1-III LIT. ________

CV

[5] PLACE 1-XSW-235-3, 120V AC VITAL INSTR POWER BD 1-III TRANSFER, on 120V AC VITAL INSTR POWER BD 1-III to ALTERNATE. ________

CV

[6] CHECK 120V AC VITAL INSTR POWER BD 1-III ENERGIZED. ________

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]:

A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)]

Some examples of SRO exam items for this topic include:

  • Reporting requirements when the maximum licensed thermal power output is exceeded.
  • The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
  • National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
  • Processes for TS and FSAR changes.

Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic.

B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

Some examples of SRO exam items for this topic include:

The question requires the

  • Application of Required Actions (Section 3) and Surveillance knowledge of the Requirements (SR) (Section 4) in accordance with rules of application basis for T/S LCO requirements (Section 1).

3.8.7 and 3.8.9.

  • Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
  • Knowledge of TS bases that are required to analyze TS required actions and terminology.

SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.

SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. ROs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Page 3 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)

RO knowledge Above this line Page 4 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)

(Tech Specs)

Can question be answered solely by knowing 1 Yes hour TS/TRM Action? RO question No Can question be answered solely by knowing the Yes LCO/TRM information listed above-the-line? RO question No Can question be answered solely by knowing the Yes TS Safety Limits? RO question No Does the question involve one or more of the following for TS, TRM, or ODCM?

  • Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
  • Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4) Yes SRO-only
  • Knowledge of TS bases that is required to analyze TS question required actions and terminology No Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only Page 5 of 16

88.

Given the following conditions:

- Unit 1 is at 100% power.

- The following equipment is INOPERABLE:

1. The TDAFWP rooms DC emergency exhaust fan.
2. The TDAFWP rooms AC emergency exhaust fan.
3. 1-LCV-3-164, MDAFWP SG 1 SUPPLY.

Of the THREE equipment items listed above, which ONE of the following lists the items that are required to be OPERABLE in accordance with T/S LCO 3.7.5, AFW System?

A. ONLY 3 B. ONLY 2 and 3 C. ONLY 1 and 2 D. ONLY 1 and 3

CORRECT ANSWER: D DISTRACTOR ANALYSIS:

A. Incorrect: As seen in WBN-SDD-N3-30AB-4001, Auxiliary Building Heating, Ventilation, Air Conditioning System, the TDAFW pump rooms are normally ventilated by the AB air exhaust system. Two 100% emergency exhaust fans, one AC operated and one DC operated, are provided in each TDAFW pump roomThe DC-operated fan is installed to provide the required cooling in the event of a loss of all AC power, and will automatically start upon the start of the TDAFW pump. This is the only AFW pump available during a loss of all onsite and offsite AC power. The DC fan is the only means available to maintain the temperature requirements in the room. The DC fan is therefore a primary safety-related system componentThe AC fan does not serve a safety-related function. Therefore it is correct that in accordance with the AB HVAC system description, the TDAFWP rooms DC emergency exhaust fan is safety-related.

The basis for T/S LCO 3.7.5 reflects that: This requires that the two motor driven AFW pumps be OPERABLE in two diverse paths, each supplying AFW to separate steam generators. Therefore, each MDAFWP must be able to provide two S/Gs with AFW (or stated, the MDAFWPs must be able to provide all of the S/Gs). If one of the LCVs is INOPERABLE, then that train of AFW is rendered INOPERABLE.

B. Incorrect: While it is correct that the LCVs INOPERABILITY renders T/S LCO 3.7.5 NOT MET, it is not correct that the AC exhaust fan serves a safety related purpose and as such is required by the T/S. It is plausible to believe that it does so as its name implies such (i.e. the AC emergency exhaust fans [emphasis added]).

C. Incorrect: While it is correct that in accordance with the AB HVAC system description, the TDAFWP rooms DC emergency exhaust fan is safety-related, it is neither correct that the AC emergency exhaust fan is required to serve a safety related function and as such is required OPERABLE by the T/S nor is it correct that the loss of a single LCV would allow the train of AFW to remain OPERABLE. It is plausible to believe such as one could reason that as long as a MDAFWP could feed one S/G that it would remain OPERABLE.

D. Correct: It is correct that both the LCV and the DC emergency exhaust fan are required OPERABLE for T/S LCO 3.7.5 to be MET.

Question Number: 88 Tier: 2 Group: 1 K/A: 061 Auxiliary / Emergency Feedwater System 2.2 Equipment Control 2.2.22 Knowledge of limiting conditions for operations and safety limits.

Importance Rating: 4.0 4.7 10 CFR Part 55: (CFR: 41.5 / 43.2 / 45.2) 10CFR55.43.b: 10 CFR 55.43(b)(2)

K/A Match: K/A is matched because the applicant is required determine which items (as provided in an equipment list) affect the OPERABILITY of the AFW system.

Technical

Reference:

Basis for T/S LCO 3.7.5, AFW System WBN-SDD-N3-30AB-4001, Auxiliary Building Heating, Ventilation, Air Conditioning System Proposed references to None be provided:

Learning Objective: 3-OT-SYS003B

13. DESCRIBE the following aspects of TS and TRs
b. The Limiting Conditions for Operation, Applicability, and Bases.

Cognitive Level:

Higher Lower X Question Source:

New X Modified Bank Bank Question History: New question for the 2015-301 NRC SRO Exam Comments:

AFW System B 3.7.5 BASES (continued)

LCO This LCO provides assurance that the AFW System will perform its design safety function to mitigate the consequences of accidents that could result in overpressurization of the reactor coolant pressure boundary. Three independent AFW pumps in three diverse trains are required to be OPERABLE to ensure the The MDAFWPs availability of RHR capability for all events accompanied by a loss of offsite power must be able to and a single failure. This is accomplished by powering two of the pumps from supply 2 S/G per independent emergency buses. The third AFW pump is powered by a different means, a steam driven turbine supplied with steam from a source that is not MDAFWP (i.e. all isolated by closure of the MSIVs.

four S/Gs must have an operable The AFW System is considered OPERABLE when the components and flow feed) paths required to provide redundant AFW flow to the steam generators are OPERABLE. This requires that the two motor driven AFW pumps be OPERABLE in two diverse paths, each supplying AFW to separate steam generators. The turbine driven AFW pump is required to be OPERABLE with redundant steam supplies from each of two main steam lines upstream of the MSIVs, and shall be capable of supplying AFW to any of the steam generators.

The piping, valves, instrumentation, and controls in the required flow paths also are required to be OPERABLE.

The LCO is modified by a Note indicating that one AFW train, which includes a motor driven pump, is required to be OPERABLE in MODE 4. This is because of the reduced heat removal requirements and short period of time in MODE 4 during which the AFW is required and the insufficient steam available in MODE 4 to power the turbine driven AFW pump.

APPLICABILITY In MODES 1, 2, and 3, the AFW System is required to be OPERABLE in the event that it is called upon to function when the MFW is lost. In addition, the AFW System is required to supply enough makeup water to replace the steam generator secondary inventory, lost as the unit cools to MODE 4 conditions.

In MODE 4 the AFW System may be used for heat removal via the steam generators.

In MODE 5 or 6, the steam generators are not normally used for heat removal, and the AFW System is not required.

(continued)

Watts Bar-Unit 1 B 3.7-27

WBN System AUXILIARY BUILDING HEATING, WBN-SDD-N3-30AB-4001 Description VENTILATION, AIR CONDITIONING Rev. 0038 Document SYSTEM (30, 31, 44) Page 63 of 201 2.1.1 Safety Function (continued)

B. Safe shutdown earthquake (SSE)

C. Loss of offsite power (LOOP)

D. Tornado E. Flood F. Airborne radioactive contamination Note: The TDAFW Pump Room DC powered exhaust fan is required to function during a loss of all AC power.

The required DBEs and associated safety functions for the system are tabulated in Ref 7.2.22.

The ABGTS and ABSCE serve a primary safety function by (1) providing a secondary containment barrier maintained under negative pressure during certain postulated accidents involving airborne radioactivity except a Fuel Handling Accident, and (2) providing contaminant removal sufficient to keep radioactivity levels in the air released to the environment low enough to assure compliance with the requirements of 10CFR100 (Ref 7.5.1).

Although the ABGTS and ABSCE are available to minimize the consequences of a Fuel Handling Accident, they are not required to function in order to meet the control room and offsite dose limits of 10CRF50.67 (Ref 7.5.15) based on the use of Regulatory Guide 1.183 (Alternate Source Terms) methodology (Ref 7.5.16).

Other portions of the AB HVAC System also serve a primary safety function by maintaining acceptable environmental conditions within the building as discussed in Ref 7.2.2 for protection of ESF mechanical and electrical equipment and controls following a design basis event.

Those portions of the AB HVAC System not serving a primary safety function (See paragraph 2.1.2) perform a secondary safety function by maintaining limited structural integrity during an earthquake to prevent interactions with primary safety components which could jeopardize primary safety functions.

Mechanical devices and associated instrumentation and controls and electrical equipment which perform a primary or secondary safety function are tabulated in references 7.1.10 and 7.1.11.

2.1.2 Normal Function During normal operations the AB HVAC System shall be designed to maintain acceptable environmental conditions as discussed in Ref 7.2.2 for equipment protection, personnel access, operation, inspection, maintenance, and testing; and to limit the release of radioactivity to the environment during all weather conditions.

WBN System AUXILIARY BUILDING HEATING, WBN-SDD-N3-30AB-4001 Description VENTILATION, AIR CONDITIONING Rev. 0038 Document SYSTEM (30, 31, 44) Page 82 of 201 3.1.3 Auxiliary Building HVAC (continued)

3. Additional Equipment Building HVAC The Unit 1 additional equipment building is served by three nonsafety-related air conditioning units. One unit provides air to the spaces on EL. 729, 740.5, and 752. A second unit provides air to EL. 763 and 775. The third unit provides air to the equipment spaces on El. 786.5. Grated floor openings provide an air path for the return air back to each unit. The Unit 2 additional equipment building is served by one nonsafety air conditioning unit which provides air to El. 729 and 763. Each of the air conditioning units is designed to maintain the temperature at approximately 92°F dry bulb and 73°F wet bulb. Condensing water is provided by the raw cooling water system.

The Additional Equipment Buildings are outside the ABSCE boundary; therefore, they are not connected to the ABGTS ventilation exhaust.

4. Turbine Driven Auxiliary Feedwater (TDAFW) Pump Room Exhaust The TDAFW pump rooms are normally ventilated by the AB Air Exhaust System.

Two 100% emergency exhaust fans, one (115 volt, 60Hz) AC operated and one (115 volt) DC operated, are provided in each TDAFW pump room. Each fan is sized to provide the required air flow in the room for the volume changes method of cooling. The fans are roof ventilator type venting into the general spaces of the Auxiliary Building. The DC-operated fan is installed to provide the required cooling in the event of a loss of all AC power, and will automatically start upon the start of the TDAFW pump. This is the only AFW pump available during a loss of all onsite and offsite AC power. The DC fan is the only means available to maintain the temperature requirements in the room. The DC fan is therefore a primary safety-related system component (Ref 7.2.1). See Table 9.5 for the design parameters of the DC fan. The AC fan does not serve a safety-related function and is Seismic Category I(L)B (Ref 7.2.1). Both fans are thermostatically controlled to automatically operate at a room temperature of greater than the setpoint.

5. Sample Room Ventilation System The sample room is ventilated by five nonsafety lab hood exhaust fans. Three fans are located on the Unit 1 side and two fans are located on the Unit 2 side.

Air enters the sample room through doors with transfer grilles and backdraft dampers. Each hood is provided with a separate exhaust fan and HEPA filter assembly. The HEPA filters located upstream from each fan have a nominal efficiency of 99.97%. A differential pressure gauge indicates the need for filter replacement. Each hood exhaust fan discharges into the General Ventilation exhaust system.

6. Main Steam Valve Vault Ventilation System The Main steam valve vault rooms (south and north) each have an independent nonsafety ventilation system consisting of two roof-mounted exhaust fans. The fans draw outside air for room cooling through a wall opening near the floor.

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]:

A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)]

Some examples of SRO exam items for this topic include:

  • Reporting requirements when the maximum licensed thermal power output is exceeded.
  • The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
  • National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
  • Processes for TS and FSAR changes.

Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic.

B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

Some examples of SRO exam items for this topic include:

  • Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
  • Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
  • Knowledge of TS bases that are required to analyze TS required actions and terminology.

SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.

SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. ROs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Page 3 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)

RO knowledge Above this line Page 4 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)

(Tech Specs)

Can question be answered solely by knowing 1 Yes hour TS/TRM Action? RO question No Can question be answered solely by knowing the Yes LCO/TRM information listed above-the-line? RO question No Can question be answered solely by knowing the Yes TS Safety Limits? RO question The question requires that the applicant consider both the information in the T/S LCO basis as well No as that contained in the Aux Bldg ventilation system description to determine the compliance with T/S LCO 3.7.5.

Does the question involve one or more of the following for TS, TRM, or ODCM?

  • Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
  • Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4) Yes SRO-only
  • Knowledge of TS bases that is required to analyze TS question required actions and terminology No Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only Page 5 of 16

89.

Given the following conditions:

- Unit 1 is at 100%.

- A buzzer is heard from behind the Shift Managers desk.

- The following is observed:

- The crew enters 0-AOI-13, section 3.2, Loss of ERCW Pump.

Which ONE of the following describes T/S LCO 3.7.8, ERCW?

In accordance with T/S LCO 3.7.8, ________.

A. a train of ERCW is INOPERABLE and will ONLY be restored OPERABLE when the failed pump is repaired and retested B. the failure does NOT impact the operability of the ERCW system and in accordance with OPDP-8, the crew will enter a TRACKING ONLY LCO for the ERCW system C. a train of ERCW is INOPERABLE and will be restored OPERABLE IMMEDIATELY after the crew performs step 1 of 0-AOI-13, START redundant trained ERCW Pump D. a train of ERCW is INOPERABLE and will be restored OPERABLE IMMEDIATELY after the crew performs step 4 of 0-AOI-13, ENSURE applicable emergency power selector switch selected away from failed pump

CORRECT ANSWER: D DISTRACTOR ANALYSIS:

A. Incorrect, T/S LCO 3.7.8, ERCW states: Two ERCW trains shall be OPERABLE. The basis for this T/S declares: An ERCW train is considered OPERABLE during MODES 1, 2, 3, and 4 when: Two pumps, aligned to separate shutdown boards, are OPERABLE. This indicates that one ERCW pump must be available per shutdown board. The basis statement Two pumps per train are aligned to receive power from different diesel generators indicates the T/S impact of the emergency power selector switch. The emergency power selector switch designates which ERCW pump will automatically start either after a blackout or a safety injection. The ERCW pump not selected will not automatically start after either a blackout or SI. The one ERCW pump per Shutdown Board required OPERABLE by the T/S is therefore, that ERCW pump which is selected by the emergency power selector switch.

In the conditions depicted in the stem of the question, the A-A ERCW pump has tripped. The buzzer, white and green indicating lights and lack of pressure indication all relate that the pump had been running and is now tripped. The emergency power selector switch can be seen in the A-A pump position. Therefore, an inoperable pump is selected by the selector switch. Therefore, it is true that T/S LCO 3.7.8 is NOT MET at the time that the indications are beheld.

It is incorrect to believe that the ERCW train will only be restored OPERABLE when the A-A ERCW pump is repaired and retested. It is plausible to believe this as one may believe that all four ERCW pumps per train are required operable by the T/S.

B. Incorrect, This distractor is incorrect as detailed above. It would be correct if the emergency power selector switch had been seen in the B-A position.

The B-A pump is OPERABLE and would be selected for emergency start. Therefore, the crew would enter a tracking-only T/S for LCO 3.7.8.

C. Incorrect, This distractor is incorrect as detailed above. It is plausible given that if the question of OPERABILITY were treated in the same manner that a system such as CCS was to be, then the applicant would arrive at this distractor. Specifically, take the example of the Unit 1, A train CCS. If the 1B CCS pump were running (aligned as normal to the A train) and then tripped with the 1A pump failing to start in automatic, T/S LCO 3.7.7, CCS would at that point be NOT MET. Subsequently, if an operator took the 1A CCS pump to start and did start the pump, then the T/S LCO 3.7.7 would be MET.

D. Correct, As detailed above, the only action required to restore the operability of the A train ERCW is to reposition the emergency power selector switch away from the failed pump.

Question Number: 89 Tier: 2 Group: 1 K/A: 076 Service Water System (SWS)

A2 Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based on those predictions, use procedures to Correct: control, or mitigate the consequences of those malfunctions or operations:

A2.01 Loss of SWS Importance Rating: 3.5 3.7 10 CFR Part 55: (CFR: 41.5 / 43.5 / 45/3 / 45/13) 10CFR55.43.b: 10 CFR 55.43(b)(2)

K/A Match: The K/A is matched because given an entry into 0-AOI-13, Loss of ERCW (Loss of SWS) the applicant must predict the impact on the OPERABILITY of ERCW (SWS) which is had. Subsequently, the applicant must use procedures (0-AOI-13 and the T/S) to mitigate the impact that the Loss of ERCW had upon the OPERABILITY of the system.

1-45W760-67-1 Technical

Reference:

0-AOI-13, Loss of Essential Raw Cooling Water T/S LCO 3.7.8, ERCW Proposed references to None be provided:

Learning Objective: 3-OT-SYS067A

12. DESCRIBE the following aspects of TS and TRs
b. The Limiting Conditions for Operation, Applicability, and Bases.

Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History:

New question for the 2015-301 NRC SRO Exam Comments:

ERCW B 3.7.8 B 3.7 PLANT SYSTEMS B 3.7.8 Essential Raw Cooling Water (ERCW) System BASES BACKGROUND The ERCW provides a heat sink for the removal of process and operating heat from safety related components during a Design Basis Accident (DBA) or transient. During normal operation, and a normal shutdown, the ERCW System also provides this function for various safety related and nonsafety related components. The safety related function is covered by this LCO.

The shared ERCW system consists of eight 50% ERCW pumps, four traveling water screens, four screen wash pumps, four strainers, associated piping, valves, and instrumentation.

Water for the ERCW system enters two separate sump areas of the pumping station through four traveling water screens, two for each sump. Four ERCW pumping units, all on the same plant train, take suction from one of the sumps, The ERCW pumps and four more on the opposite plant train take suction from the other sump. One set of pumps and associated equipment is designated Train A, and the other which will start Train B. These trains are redundant and are normally maintained separate and upon the receipt of independent of each other. Each set of four pumps discharges into a common the SI signal are manifold, from which two separate headers (1A and 2A for Train A, and 1B and those selected by 2B for Train B) each with its own automatic backwashing strainer, supply water to the ERCW PMP the various system users. Two pumps per train are adequate to supply worst case conditions. Two pumps per train are aligned to receive power from different DG POWER SEL diesel generators. Operator designated pumps and valves are remote and switches. manually aligned, except in the unlikely event of a loss-of-coolant accident (LOCA). The pumps are automatically started upon receipt of a safety injection (SI) signal, and some essential valves are aligned to their post-accident positions.

Some manual realignments of motor-operated valves (MOVs) are necessary.

The ERCW System also provides emergency makeup to the Component Cooling System (CCS) and is the backup water supply to the Auxiliary Feedwater System.

Additional information about the design and operation of the ERCW, along with a list of the components served, is presented in the FSAR, Section 9.2.1 (continued)

Watts Bar-Unit 1 B 3.7-43

ERCW B 3.7.8 BASES BACKGROUND (Ref. 1). The principal safety related function of the ERCW System is the (continued) removal of decay heat from the reactor via the CCS.

APPLICABLE The design basis of the ERCW System is for one ERCW train, in conjunction SAFETY ANALYSES with the CCS and a 100% capacity Containment Spray System and Residual Heat Removal (RHR), to remove core decay heat following a design basis LOCA as discussed in the FSAR, Section 9.2.1 (Ref. 1). This prevents the containment sump fluid from increasing in temperature during the recirculation phase following a LOCA and provides for a gradual reduction in the temperature of this fluid as it is supplied to the Reactor Coolant System by the ECCS pumps. The ERCW System is designed to perform its function with a single failure of any active component, assuming the loss of offsite power.

The ERCW System, in conjunction with the CCS, also cools the unit from RHR, as discussed in the FSAR, Section 5.5.7, (Ref. 2) entry conditions to MODE 5 during normal and post accident operations. The time required for this evolution is a function of the number of CCS and RHR System trains that are operating.

One ERCW train is sufficient to remove decay heat during subsequent operations in MODES 5 and 6. This assumes a maximum ERCW temperature of 85 F occurring simultaneously with maximum heat loads on the system.

The ERCW System satisfies Criterion 3 of the NRC Policy Statement.

LCO Two ERCW trains are required to be OPERABLE to provide the required redundancy to ensure that the system functions to remove post accident heat loads, assuming that the worst case single active failure occurs coincident with the loss of offsite power.

An ERCW train is considered OPERABLE during MODES 1, 2, 3, and 4 when:

(continued)

Watts Bar-Unit 1 B 3.7-44

The aligned to a SDBD means that ERCW the ERCW PMP B 3.7.8 DG POWER SEL BASESswitch is positioned to the pump.

LCO a. Two pumps, aligned to separate shutdown boards, are OPERABLE; and (continued)

b. The associated piping, valves, heat exchanger, and instrumentation and controls required to perform the safety related function are OPERABLE.

APPLICABILITY In MODES 1, 2, 3, and 4, the ERCW System is a normally operating system that is required to support the OPERABILITY of the equipment serviced by the ERCW System and required to be OPERABLE in these MODES.

In MODES 5 and 6, the OPERABILITY requirements of the ERCW System are determined by the systems it supports.

ACTIONS A.1 If one ERCW train is inoperable, action must be taken to restore OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In this Condition, the remaining OPERABLE ERCW train is adequate to perform the heat removal function. However, the overall reliability is reduced because a single failure in the OPERABLE ERCW train could result in loss of ERCW System function. Required Action A.1 is modified by two Notes.

The first Note indicates that the applicable Conditions and Required Actions of LCO 3.8.1, "AC Sources Operating," should be entered if an inoperable ERCW train results in an inoperable emergency diesel generator. The second Note indicates that the applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops MODE 4," should be entered if an inoperable ERCW train results in an inoperable decay heat removal train. This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is based on the redundant capabilities afforded by the OPERABLE train, and the low probability of a DBA occurring during this time period.

B.1 and B.2 If the ERCW train cannot be restored to OPERABLE status within the associated Completion Time, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least (continued)

Watts Bar-Unit 1 B 3.7-45

WBN Loss of Essential Raw Cooling Water 0-AOI-13 Unit 0 Rev. 0001 3.0 OPERATOR ACTIONS 3.1 Diagnostics IF GO TO PAGE SECTION Loss of ERCW pump or indications of broken pump shaft: 3.2 6 Motor trip out alarm OR Low amps and discharge pressure on running pump Supply Header Rupture in Auxiliary Building; 3.3 8 HIGH flow on supply header AND Building flood alarm LIT.

Supply Header Rupture in Yard/Downstream of Strainer: 3.4 17 Strainer DP alarm LIT AND LOW flow on individual supply header with LOW pressure on IPS supply header.

If IPS strainer room sump alarm is LIT rupture may be downstream of strainer in strainer room.

Plugged Strainer: 3.4 17 Strainer DP alarm LIT AND LOW flow on individual supply header with HIGH pressure indicated on IPS supply header.

Supply Header Rupture in IPS; 3.5 28 Supply headers flow LOW AND IPS header pressure LOW with Strainer DP alarm DARK, AND IPS strainer room sump alarm LIT.

Discharge Header Rupture in Auxiliary Building: 3.6 36 Building flood alarm LIT AND Supply header flows NORMAL.

Loss of flow on ALL ERCW supply headers 3.7 42 Page 5 of 88

WBN Loss of Essential Raw Cooling Water 0-AOI-13 Unit 0 Rev. 0001 Step Action/Expected Response Response Not Obtained 3.2 Loss of ERCW Pump

1. CHECK header pressure and flows START redundant trained ERCW adequate for current conditions. Pump.
2. ENSURE pump amps NORMAL.
3. PLACE failed pump HS in PULL TO LOCK.

Notice that T/S

4. ENSURE applicable emergency LCO 3.7.8 is NOT power selector switch selected away met from the point from failed pump. of pump loss until the completion of step 4.
5. DISPATCH personnel to determine reason for pump failure.
6. ENSURE header pressures and flows IF ERCW header pressures and flows return to expected values for existing cannot be returned to NORMAL, THEN plant conditions. **GO TO Section 3.1 Diagnostics to evaluate for a potential rupture.

Page 6 of 88

WBN Loss of Essential Raw Cooling Water 0-AOI-13 Unit 0 Rev. 0001 Step Action/Expected Response Response Not Obtained 3.2 Loss of ERCW Pump (continued)

7. CLOSE discharge valve on failed pump.

A TRAIN DISCHARGE VALVE B TRAIN DISCHARGE VALVE PUMPS PUMPS A-A 0-ISV-67-504A E-B 0-ISV-67-504E B-A 0-ISV-67-504B F-B 0-ISV-67-504F C-A 0-ISV-67-504C G-B 0-ISV-67-504G D-A 0-ISV-67-504D H-B 0-ISV-67-504H

8. INITIATE repair.
9. REFER TO Tech Spec 3.7.8, Again, T/S LCO Essential Raw Cooling Water System 3.7.8 was not met (ERCW). for the mentioned period. After such,
10. RETURN TO Instruction in effect. a TRACKING ONLY LCO is End of Section entered.

Page 7 of 88

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]:

A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)]

Some examples of SRO exam items for this topic include:

  • Reporting requirements when the maximum licensed thermal power output is exceeded.
  • The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
  • National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
  • Processes for TS and FSAR changes.

Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic.

B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

Some examples of SRO exam items for this topic include:

  • Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
  • Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
  • Knowledge of TS bases that are required to analyze TS required actions and terminology.

SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.

SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. ROs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Page 3 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)

RO knowledge Above this line Page 4 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)

(Tech Specs)

Can question be answered solely by knowing 1 Yes hour TS/TRM Action? RO question No Can question be answered solely by knowing the Yes LCO/TRM information listed above-the-line? RO question No Can question be answered solely by knowing the Yes TS Safety Limits? RO question No Does the question involve one or more of the following for TS, TRM, or ODCM?

  • Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
  • Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4) Yes SRO-only
  • Knowledge of TS bases that is required to analyze TS question required actions and terminology No Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only Page 5 of 16

90.

Given the following conditions:

- Unit 1 is at 100% power.

- Containment Pressure Transmitter 1-PDT-30-43 (Channel III) FAILED and is out of service with the channel bistables positioned as required by Tech Specs.

- The Surveillance Instruction for 1-PDT-30-44 (Channel II) is NOW due.

Which ONE of the following describes the required action for performing the Surveillance Instruction on 1-PDT-30-44 and the impact on Containment Spray actuation?

1-PDT-30-43 is required to be placed in the ____(1)____ position AND subsequent testing of 1-PDT-30-44 will ____(2)____ a valid AUTOMATIC Containment Spray actuation.

A. (1) BYPASS (2) ALLOW B. (1) BYPASS (2) PREVENT C. (1) TRIPPED (2) ALLOW D. (1) TRIPPED (2) PREVENT

CORRECT ANSWER: A DISTRACTOR ANALYSIS:

A. Correct, 1-PDT-30-43 would be placed in the bypass position as identified in the Tech Spec 3.3.2 Bases. The Required Action for LCO 3.3.2 Condition E has a Note that allows a channel to be bypassed for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for surveillance testing. This Note is explained in the Tech Spec Bases. There are 4 channels provided for the Hi-Hi containment function to actuate and it takes 2 of the 4 to generate the signal (and 2 channels remain in service).

B. Incorrect, 1-PDT-30-43 would be placed in the bypass position but the testing of 1-PDT-30-44 will not prevent a valid containment spray actuation from occurring even though the HI-HI bistables would be tested in the bypass position.

C. Incorrect, 1-PDT-30-43 will not be placed to the trip position and subsequent testing of 1-PDT-30-44 will still allow valid automatic actuation of the containment spray during the surveillance test because the other 2 channels still provide for the protection of the function. Plausible because the positioning of bistables to the trip position is a normal process performed to meet Tech Spec requirements to ensure a function can be met and in this COT the Hi containment pressure SI bistable for this instrument would be placed in the Trip position unless another channel was to be tested.

D. Incorrect, 1-PDT-30-43 will not be placed to the trip position and subsequent testing of 1-PDT-30-44 will not prevent a valid automatic actuation of the containment spray during the surveillance test because the other 2 channels still provide for the protection of the function. Plausible because the positioning of bistables to the trip position is a normal process performed to meet Tech Spec requirements to ensure a function can be met (and in this COT the Hi containment pressure SI bistable for this instrument would be placed in the Trip position unless testing of another channel was required.

Question Number: 90 Tier: 2 Group: 1 K/A: 026 Containment Spray System A2 Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; and (b) based on those predictions, use procedures to Correct: control, or mitigate the consequences of those malfunctions or operations:

A2.03 Failure of ESF Importance Rating: 4.1 4.4 10 CFR Part 55: (CFR: 41.5 / 43.5 / 45.3 / 45.13) 10CFR55.43.b: 10 CFR 55.43(b)(2)

K/A Match: Applicant must determine how the bistables will be concurrently configured on a failed Containment Spray System actuation ESF transmitter and an ESF transmitter which is required to be tested in order to run the surveillance instruction and how the function is maintained as identified in the Technical Specification bases.

Technical

Reference:

T/S LCO 3.3.2, ESFAS Instrumentation T/S Basis for LCO 3.3.2 1-47W611-88-1 Proposed references to None be provided:

Learning Objective: 3-OT-SYS072A

11. DESCRIBE the following aspects of TS and TRs
b. The Limiting Conditions for Operation, Applicability, and Bases.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: Bank question 026A2.03 88. Used on the 11/2009 WBN NRC exam.

Comments:

ESFAS Instrumentation 3.3.2 3.3 INSTRUMENTATION 3.3.2 Engineered Safety Feature Actuation System (ESFAS) Instrumentation LCO 3.3.2 The ESFAS instrumentation for each Function in Table 3.3.2-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.2-1.

ACTIONS

NOTE-------------------------------------

Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions with one A.1 Enter the Condition referenced Immediately or more required channels or in Table 3.3.2-1 for the trains inoperable. channel(s) or train(s).

B. One channel or train inoperable. B.1 Restore channel or train to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> OPERABLE status.

OR B.2.1 Be in MODE 3. 54 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br /> AND B.2.2 Be in MODE 5. 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> (continued)

Watts Bar-Unit 1 3.3-24

ESFAS Instrumentation 3.3.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. One Containment E.1 ---------------NOTE----------------

Pressure channel inoperable. One channel may be bypassed for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for surveillance testing.

Place channel in bypass. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OR E.2.1 Be in MODE 3. 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> AND E.2.2 Be in MODE 4. 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> F. One channel or train F.1 Restore channel or 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> inoperable. train to OPERABLE status.

OR F.2.1 Be in MODE 3. 54 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br /> AND F.2.2 Be in MODE 4. 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> (continued)

Watts Bar-Unit 1 3.3-26 Amendment 68

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 1 of 7)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES NOMINAL OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

1. Safety Injection
a. Manual 1, 2, 3, 4 2 B SR 3.3.2.8 NA NA Initiation
b. Automatic 1, 2, 3, 4 2 trains C SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.3 and Actuation SR 3.3.2.5 Relays SR 3.3.2.7
c. Containment 1, 2, 3 3 D SR 3.3.2.1 1.6 psig 1.5 psig Pressure- SR 3.3.2.4 High SR 3.3.2.9 SR 3.3.2.10 (a)
d. Pressurizer 1, 2, 3 3 D SR 3.3.2.1 1864.8 psig 1870 psig Pressure-Low SR 3.3.2.4 SR 3.3.2.9 SR 3.3.2.10 HI pressure goes
e. Steam Line 1, 2, 3 (a) 3 per steam D SR 3.3.2.1 to666.6 trip.(b) psig (b) 675 psig Pressure-Low line SR 3.3.2.4 SR 3.3.2.9 SR 3.3.2.10 HI HI pressure
2. Containment Spray goes to bypass
a. Manual 1, 2, 3, 4 2 per train, B SR 3.3.2.8 (theNAtwo are NA Initiation 2 trains different).
b. Automatic 1, 2, 3, 4 2 trains C SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.3 and Actuation SR 3.3.2.5 Relays
c. Containment 1, 2, 3 4 E SR 3.3.2.1 2.9 psig 2.8 psig Pressure- SR 3.3.2.4 High High SR 3.3.2.9 SR 3.3.2.10 (continued)

(a) Above the P-11 (Pressurizer Pressure) Interlock.

(b) Time constants used in the lead/lag controller are t1 50 seconds and t2 5 seconds.

Watts Bar-Unit 1 3.3-34

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 2 of 7)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES NOMINAL OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

3. Containment Isolation
a. Phase A Isolation (1) Manual 1, 2, 3, 4 2 B SR 3.3.2.8 NA NA Initiation (2) Automatic 1, 2, 3, 4 2 trains C SR 3.3.2.2 NA NA Actuation SR 3.3.2.3 Logic and SR 3.3.2.5 Actuation SR 3.3.2.7 Relays (3) Safety Refer to Function 1 (Safety Injection) for all initiation Injection functions and requirements.
b. Phase B Isolation (1) Manual 1, 2, 3, 4 2 per train, B SR 3.3.2.8 NA NA Initiation 2 trains (2) Automatic 1, 2, 3, 4 2 trains C SR 3.3.2.2 NA NA Actuation SR 3.3.2.3 Logic and SR 3.3.2.5 Actuation SR 3.3.2.7 Relays (3) Con- 1, 2, 3 4 E SR 3.3.2.1 2.9 psig 2.8 psig tainment SR 3.3.2.4 Pressure-- SR 3.3.2.9 High High SR 3.3.2.10
4. Steam Line Isolation (c) (c)
a. Manual 1, 2 , 3 1/valve F SR 3.3.2.8 NA NA Initiation (c) (c)
b. Automatic 1, 2 , 3 2 trains G SR 3.3.2.2 NA NA Actuation SR 3.3.2.3 Logic and SR 3.3.2.5 Actuation Relays (continued)

(c) Except when all MSIVs are closed and de-activated.

Watts Bar-Unit 1 3.3-35

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 3 of 7)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES NOMINAL OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

4. Steam Line Isolation (continued)

(c) (c)

c. Containment 1, 2 , 3 4 E SR 3.3.2.1 2.9 psig 2.8 psig Pressure- SR 3.3.2.4 High High SR 3.3.2.9 SR 3.3.2.10
d. Steam Line Pressure (c)

(1) Low 1, 2 , 3(a) (c) 3 per steam D SR 3.3.2.1 666.6 (b) psig 675 (b) psig line SR 3.3.2.4 SR 3.3.2.9 SR .3.2.10 (d) (c) (e) (e)

(2) Negative 3 3 per steam D SR 3.3.2.1 108.5 psi 100 psi Rate-High line SR 3.3.2.4 SR 3.3.2.9 SR 3.3.2.10

5. Turbine Trip and Feedwater Isolation (f)
a. Automatic 1, 2 , 3(f) 2 trains H SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.3 and Actuation SR 3.3.2.5 Relays (f)
b. SG Water 1, 2 , 3(f) 3 per SG I SR 3.3.2.1 83.1% 82.4%

Level-High SR 3.3.2.4 High(P-14) SR 3.3.2.9 (h)

SR 3.3.2.10

c. Safety Refer to Function 1 (Safety Injection) for all initiation Injection functions and requirements.

(f) (g)

d. North MSV Vault 1, 2 , 3/vault O SR 3.3.2.6 5.31 inches 4 inches Room Water Room SR 3.3.2.9 Level - High (f), (g)
e. South MSV Vault 1, 2 3/vault O SR 3.3.2.6 4.56 inches 4 inches Room Water Room SR 3.3.2.9 Level - High (continued)

(a) Above the P-11 (Pressurizer Pressure) interlock.

(b) Time constants used in the lead/lag controller are t1 50 seconds and t2 5 seconds.

(c) Except when all MSIVs are closed and de-activated.

(d) Function automatically blocked above P-11 (Pressurizer Interlock) setpoint and is enabled below P-11 when safety injection on Steam Line Pressure Low is manually blocked.

(e) Time constants utilized in the rate/lag controller are t3 and t4 50 seconds.

(f) Except when all MFIVs, MFRVs, and associated bypass valves are closed and de-activated or isolated by a closed manual valve.

(g) MODE 2 if Turbine Driven Main Feed Pumps are operating.

(h) For the time period between February 23, 2000, and prior to turbine restart (following the next time the turbine is removed from service), the response time test requirement of SR 3.3.2.10 is not applicable for 1-FSV-47-027.

Watts Bar-Unit 1 3.3-36 Amendment 23

ESFAS Instrumentation B 3.3.2 BASES ACTIONS D.1, D.2.1, and D.2.2 (continued)

Failure to restore the inoperable channel to OPERABLE status or place it in the tripped condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> requires the plant be placed in MODE 3 within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 4 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODE 4, these Functions are no longer required OPERABLE.

The Required Actions have been modified by a Note that allows placing an inoperable channel in the bypassed condition for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> while performing routine surveillance testing of other channels. The Note also allows a channel to be placed in bypass for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for testing of the bypassed channel.

However, only one channel may be placed in bypass at any one time. The 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed for testing are justified in Reference 17.

E.1, E.2.1, and E.2.2 Condition E applies to:

Containment Spray Containment Pressure-High High; Steam Line Isolation Containment Pressure-High High; and Containment Phase B Isolation Containment Pressure-High High.

None of these signals has input to a control function. Thus, two-out-of-three logic is necessary to meet acceptable protective requirements. However, a two-out-of-three design would require tripping a failed channel. This is undesirable because a single failure would then cause spurious containment spray initiation. Spurious spray actuation is undesirable because of the cleanup problems presented. Therefore, these channels are designed with (continued)

Watts Bar-Unit 1 B 3.3-104 Revision 90 Amendment 68

ESFAS Instrumentation B 3.3.2 BASES ACTIONS E.1, E.2.1, and E.2.2 (continued) two-out-of-four logic so that a failed channel may be bypassed rather than tripped. Note that one channel may be bypassed and still satisfy the single failure criterion. Furthermore, with one channel bypassed, a single instrumentation channel failure will not spuriously initiate containment spray.

To avoid the inadvertent actuation of containment spray and Phase B containment isolation, the inoperable channel should not be placed in the tripped condition. Instead it is bypassed. Restoring the channel to OPERABLE status, or placing the inoperable channel in the bypass condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, is sufficient to assure that the Function remains OPERABLE and minimizes the time that the Function may be in a partial trip condition (assuming the inoperable channel has failed high). The Completion Time is further justified based on the low probability of an event occurring during this interval. Failure to restore the inoperable channel to OPERABLE status, or place it in the bypassed condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, requires the plant be placed in MODE 3 within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 4 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODE 4, these Functions are no longer required OPERABLE.

The Required Actions are modified by a Note that allows placing one channel in bypass for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> while performing routine surveillance testing. The channel to be tested can be tested in bypass with the inoperable channel also in bypass. The time limit is justified in Reference 17.

F.1, F.2.1, and F.2.2 Condition F applies to:

Manual Initiation of Steam Line Isolation; Loss of Offsite Power; Auxiliary Feedwater Pump Suction Transfer on Suction PressureLow; and (continued)

Watts Bar-Unit 1 B 3.3-105 Revision 90 Amendment 68

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]:

A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)]

Some examples of SRO exam items for this topic include:

  • Reporting requirements when the maximum licensed thermal power output is exceeded.
  • The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
  • National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
  • Processes for TS and FSAR changes.

Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic.

B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

Some examples of SRO exam items for this topic include:

  • Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
  • Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
  • Knowledge of TS bases that are required to analyze TS required actions and terminology.

SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.

SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. ROs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Page 3 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)

The knowledge required to complete this question is "below the line."

RO knowledge Above this line Page 4 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)

(Tech Specs)

Can question be answered solely by knowing 1 Yes hour TS/TRM Action? RO question No Can question be answered solely by knowing the Yes LCO/TRM information listed above-the-line? RO question No Can question be answered solely by knowing the Yes TS Safety Limits? RO question No Does the question involve one or more of the following for TS, TRM, or ODCM?

  • Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
  • Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4) Yes SRO-only
  • Knowledge of TS bases that is required to analyze TS question required actions and terminology No Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only Page 5 of 16

91.

Given the following timeline:

00:00:00 - Unit 1 is at 100% power.

00:0 1 :00 - 1-FR-S.1, Nuclear Power Generation/ATWS is entered.

00: 1 5:00 - In accordance with step 19, the crew CHECKS Incore T/Cs.

- ALL T/Cs are greater than (>) 1200°F slowly RISING .

- ALL Power Range Detectors indicate 4%.

- ALL Intermediate Range SUR Monitors indicate -0.1 dpm.

Which ONE of the following completes the statements below?

At 00: 1 5:00, T (T = twall - tcc), as shown in the picture above, is ____(1)____ it was at 00:00:00.

At 00: 1 5:0 1, in accordance with 1-FR-S.1, the crew will be required to GO TO _____(2)_____.

A. (1) the same as (2) SACRG-1, Severe Accident Control Room Guideline Initial Response B. (1) greater than (2) SACRG-1, Severe Accident Control Room Guideline Initial Response C. (1) the same as (2) 1-FR-C.1, Inadequate Core Cooling D. (1) greater than (2) 1-FR-C.1, Inadequate Core Cooling

CORRECT ANSWER: B DISTRACTOR ANALYSIS:

A. Incorrect: WCAP-9753, Inadequate Core Cooling Studies of Scenarios With Feedwater Available, Using the NOTRUMP Computer Code, was produced by Westinghouse and demonstrated that while the bulk fluid temperature (as measured by the incore thermocouples) was not equal to the fuel metal temperature it was related to such.

As seen in the Westinghouse background document for the status tree F-0.2, Core Cooling. the 1200°F setpoint for incore temperature indicates that most liquid inventory has already been removed from the RCS and that core decay heat is superheating steam in the core.

The study, Thermal-hydraulic behavior of electrically heated [rod] during a critical heat flux transient, demonstrates that once the heat flux present within a process flowpath increases such that nucleate boiling can no longer exist, the heat transfer coefficient between the electrically heated rod (e.g. nuclear fuel rod) and the process flow (e.g. Reactor Coolant System fluid) drops dramatically. Also demonstrated (and thus expected) is the fact that the surface temperature of the electrically heated rod rises incredibly (grossly in excess of that of the bulk-albeit superheated-coolant).

WCAP-9753 further demonstrated for given cases of depleted coolant inventories that when the incore T/Cs read 1200°F, the fuel metal temperature reached temperatures over 2000 °F. This indicated a T (between the wall of the clad and the bulk coolant) of over 800°F. Depending upon which fuel location was chosen-the temperature difference could be much higher.

By contrast, at 100% power, section 4.4.2.2.5 of the UFSAR states that, the outer surface of the fuel rod at the hot spot operates at a temperature of approximately 660°F for steady state operation at rated power throughout core life due to the onset of nucleate boiling. As can be found on the integrated computer system or on the RVLIS thermocouple displays themselves, the average of all of the incore thermocouples is approximately 620°F. Therefore, the nominal difference between the hottest fuel rods metal temperature and that registered by the incore thermocouples is approximately 40°F at 100% power.

Therefore, it is incorrect to believe that the T specified in the stem of the question remains the same between the two times. It is plausible to believe this given a failure to understand the heat transfer properties in effect for the fuel during this accident.

Step 19 of 1-FR-S.1 is as follows:

19. CHECK Incore T/Cs less than 1200°F.

The response not obtained for step 19 is:

IF Incore T/Cs are greater than 1200°F AND rising, THEN

    • GO TO 1-SACRG-1, Severe Accident Control Room Guideline Initial Response.

Step 20 is:

20. CHECK reactor subcritical:
a. Power range channels less than 5%.
b. Intermediate range startup rate NEGATIVE.

Contained within the step 20 response not obtained is:

IF red OR orange condition exists on other Status Trees, THEN PERFORM actions of other FR Procedures which do not cool down or otherwise add positive reactivity to the core.

Therefore, it is correct that a transition to 1-SACRG-1 is warranted.

B. Correct: It is correct that at 00:15:00, the T specified in the stem of the question is greater than it was at 00:00:00. Also, it is correct to transition to 1-SACRG-1.

C. Incorrect: Again, it is incorrect that the T specified in the stem of the question remains the same. Also, it is incorrect that a transition to 1-FR-C.1 would be required. It is plausible to believe this as if one misses the transition to 1-SACRG-1; one will arrive at step 20 with the conditions met for a RED path to 1-FR-C.1. One could utilize the response not obtained for step 20 to perform actions of other FR procedures (in this case 1-FR-C.1).

D. Incorrect: While it is correct that at 00:15:00, the T specified in the stem of the question is greater than it was at 00:00:00. It is not correct that a transition to 1-FR-C.1 would be required.

Question Number: 91 Tier: 2 Group: 2 K/A: 017 In-Core Temperature Monitor (ITM) System A2 Ability to (a) predict the impacts of the following malfunctions or operations on the ITM system; and (b) based on those predictions, use procedures to Correct: control or mitigate the consequences of those malfunctions or operations:

A2.02 Core damage Importance Rating: 3.6 4.1 10 CFR Part 55: (CFR: 41.5 / 43.5 / 45.3 / 45.5) 10CFR55.43.b: 10 CFR 55.43(b)(5)

K/A Match: The K/A is matched because given a case of core damage (the conditions met for entry into 1-SACRG-1), the applicant must understand the relation of the temperature indicated by the ITM to that of the fuel metal. Next, the applicant must correctly implement the functional restoration procedures to correctly transition to 1-SACRG-1 in order that the core damage (and high temperatures of ITM) be mitigated.

Technical

Reference:

Section 4.4.2.2.5 of the UFSAR ICS screen shot showing the incore T/C temperatures at 100% power WCAP-9753, Inadequate Core Cooling Studies of Scenarios With Feedwater Available, Using the NOTRUMP Computer Code.

Thermal-hydraulic behavior of electrically heated [rod] during a critical heat flux transient" Westinghouse owners group background document for status tree F-0.2, Core Cooling" 1-FR-S.1, Nuclear Power Generation/ATWS" Proposed references to None be provided:

Learning Objective: 3-OT-FRS0001

9. Given a set of plant conditions, diagnose and implement action steps, RNOs, notes and cautions of 1-FR-S.1 Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question for the 2015-301 NRC SRO Exam Comments: See the attached marked up copy of the SRO ONLY guidance.

WESTINGHOUSE PROPRIETARY CLASS 2 feedwater. The two effects combine to sharply depressurize the secon-dary (Figures 86 and 87) until, at 600 psia secondary pressure, the steam dump system is automatically isolated. These events occurring on the secondary side do not noticeably alter conditions in the primary loops, since the heat transfer coefficient between the steam generator tubes and the superheated steam on the primary side is relatively small, and the secondary to primary heat transfer is low.

At 1700 seconds when the maximum average fuel temperature of about 2200°F is attained (Figure 107), primary pressure drops below the low head safety injection shut-off head, initiating core recovery. Although the core recovers in this case with minimal operator actions, higher peak temperatures would be expected than are indicated by this analysis, since only the average core assembly is modeled and zirc-water reactions are neg 1ected.

3. Core Exit Thermocouple Response In the one inch break transient, considerable recirculation was found to occur between the upper core node, the upper plenum and the guide

~ubes. Three dimensional considerations, however, lead to the conclu-sion that the core exit thermocouples would indicate the core exit tem-perature under inadequate core cooling conditions.

For the four inch break case, however, since all core flows are upward and out the top of the core, the upper core node temperature would be expected to give a good estimate of the thermocouple reading. Note that at 1363 seconds, the upper core node temperature reaches 1200°F, before the core has completely uncovered.

4. Hot Leg Temperature Response Figures 102 and 103 present the response of the resistance thermometers in the broken and intact loop hot legs, respectively. While the broken loop hot leg temperature shows only a slight rise at about the time the 17

BLOCK DESCRIPTION TABLE FOR STATUS TREE F-0.2 BLOCK DECISION: Core Exit TCs Less Than 1200°F PURPOSE: To determine if inadequate core cooling has been reached BASIS:

Analyses of inadequate core cooling scenarios (References 1 and 2) show that core exit temperature greater than 1200°F is a satisfactory criterion for basing extreme operator action. At least 5 thermocouples should be reading greater than 1200°F. Five has been chosen to allow for thermocouples failing high. This temperature indicates that most liquid inventory has already been removed from the RCS and that core decay heat is superheating steam in the core. An extreme challenge to the fuel matrix/clad barrier is imminent and a RED priority is warranted. The appropriate guideline for functional response is FR-C.1, RESPONSE TO INADEQUATE CORE COOLING. If core exit thermocouples are less than 1200°F, then subsequent blocks check for other extreme, severe, not satisfied or satisfied conditions for the safety function.

INSTRUMENTATION:

Decay heat removal still exists Core exit thermocouples temperature indication Core damage will (even if it is a very result if this poor form of such).

KNOWLEDGE: "extreme challenge" is not resolved (i.e. via N/A implementation of 1-FR-C.1).

PLANT-SPECIFIC INFORMATION:

o The following criteria should be used to determine which thermocouples to monitor:

1) At least one thermocouple should be located as close as possible to the geometric center of the core.
2) The other thermocouples should be located at least one per quadrant over the highest power assemblies in each quadrant. The outer two rows of assemblies should be excluded, since they can receive significant cooling from steam generator drainage due to refluxing. The thermocouples should be selected at each refueling to ensure that the highest power assemblies are always being used.

F-0.2 Background 6 HP/LP-Rev. 2, 4/30/2005 HF02BG.doc

Thermalhydraulic Behavior of Electrically Heated Rod during a Critical Heat Flux Transient Rita de Cássia Fernandes de Lima Mechanical Engineering Department. Universidade Federal de Pernambuco Av. Acad. Hélio Ramos, s/n 50740-530 Recife. PE. Brazil ritalima@npd.ufpe.br Pedro Carajilescov Mechanical Engineering Department. Universidade Federal Fluminense Rua Passo da Pátria,156 24210-240 Niterói. RJ. Brazil pedroc@caa.uff.br In nuclear reactors, the occurrence of critical heat flux leads to fuel rod overheating with clad fusion and radioactive products leakage. To predict the effects of such phenomenon, experiments are performed using electrically heated rods to simulate operational and accidental conditions of nuclear fuel rods. In the present work, it is performed a theoretical analysis of the drying and rewetting front propagation during a critical heat flux experiment, starting with the application of an electrical power step from steady state condition. After the occurrence of critical heat flux, the drying front propagation is predicted. After a few seconds, a power cut is considered and the rewetting front behavior is analytically observed. Studies performed with various values of coolant mass flow rate show that this variable has more influence on the drying front velocity than on the rewetting one.

Keywords: Critical heat flux, rewetting front, drying front, thermalhydraulics, numerical simulation Introduction The power generation of a nuclear reactor is limited by the coolant capability of removing the heat generated inside the fuel rods. In PWR type reactor, this capability is determined by the occurrence of critical heat flux (CHF),

also called by DNB (Departure from Nucleated Boiling). Fuel rods overheating, due to the occurrence of a boiling crisis, during a power transient, can yield clad fusion with radioactive products leakage to the coolant. To predict this type of phenomenon, it is a common practice to perform simulations of the reactor operational and transient conditions in thermalhydraulic loops, utilizing electrically heated rods. Such simulations represent an important aspect of the reactor safety analysis.

When a critical heat flux occurs, the heated surface is covered by a vapor blanket, which will spread over the rod length with a so called drying front propagation velocity. Considering that vapor has a very low thermal conductivity, the local heat transfer coefficient is drastically reduced, provoking very high local temperatures. When the power is turned off, the wetting of the surface is re-established with a rewetting front propagation velocity. Since those electrically heated rods are very expensive, during a DNB experiment, it is necessary to turn the power off very quickly in order not to damage the rods. The prediction of the drying and rewetting front propagation, for a given experiment, can be used to establish the amount of time available for the experiment without rod damage. Several authors have studied this problem. Gunnerson & Yackle (1981) establish the difference between quench and rewetting. Yu et alii (1977) analyze the quench process on hot surfaces using bidimensional conduction. The work treats the subcooled and saturated rewetting for pressures from 1 to 69 bar. Olek et alii (1988) study the rewetting in descendent films. They consider the problem as a heat conjugated problem. Carlson (1989) analyses the expansion of the CHF region in direct heated rods. He takes into consideration also the thermocouple locations and the mechanism through which the drying front is detected by them.

The present work, analyzes the thermal behavior of a typical electrically heated rod with indirect heating, as shown in Figure 1, during a step transient of the electrical power. From a steady state condition, at a given power, it is imposed a 10% power step of its initial value in order to produce the CHF. It is, then, observed the effect of pressure and flow rate in the drying and rewetting front propagation velocities.

Theoretical Model Consider the universally adopted test section, with indirect heating, shown schematically in Figure 1. Heat is generated in the electrical resistance by an electrical current, is conducted axially and radially, and is removed by the water flowing longitudinally along the rod. The heat transfer coefficient is a function of the local coolant conditions.

In order to analyze the thermal behavior of this system, it was considered:

  • circular symmetry of the electrical heaters;
  • uniform axial heat generation;
  • no heat losses through the ends of the rod;
  • constant thermal properties of the materials of the rod;
  • homogeneous model for the water two-phase flow.

For the several regions of the rod, the heat conduction equation can be written as:

where T is the rod temperature; k, the thermal conductivity; , the density and cp , the specific heat of each material. The volumetric heat generation, q , is zero for all material except for the electrical resistance.

According to Silva Neto et alii(1983), the lumped form of the energy equation for the water coolant can be written as:

where q" is the heat flux received by the coolant; A is the cross section area of the channel; p is the rod perimeter and hf , f and G are the enthalpy, density and mass flux of the coolant, respectively.

The coupling between the rod and the flow is established by the surface heat removal given by:

where kclad is the thermal conductivity of the cladding; href is the heat transfer coefficient between cladding and coolant; Tf is the coolant temperature and Rext =R4, the outer radius of the cladding.

The correlations presented in Appendix A were considered for the heat transfer coefficient, taking into account the several heat transfer regimes.

Although several different correlations can be found in literature, these were considered adequate for the present situation.

The heat conduction equation for the rod was solved by the finite control volume method with an implicit formulation. For the water enthalpy, equations were solved iteratively.

To take into account the surface-water coupling, in any length of the coolant channel, the sequence shown in Figure 2, with the variables Twall, href and q" defined in Appendix A, was adopted.

Results A computational program was developed to analyze the critical heat flux for a several types of transients. Here the classical case of a step power transient is presented. At t=0s, the variable qo (the linear power density) is increased by 10% from its steady state value of 16 600 W/m, in order to reach the CHF. The new level is then maintained for 4.0 s and then the electric power is cut off. The entire transient lasts 4.5 s.

The tables below show the physical and geometric parameters used. The electrically heated rod is composed of a Ni-Cr ( 18 to 20% Cr and 8 to 12%

Ni ) resistance, a MgO electric insulator and stainless steel ( type 349 )

cladding. The time step used in discretization is 5 x 10 -2 s. The numbers of nodes considered are: five in the internal insulator, nine in the resistance, four in the internal insulator and four in the cladding. The axial interval is equal to 10-2 m.

The variation of the heat transfer coefficient versus height in the coolant channel, for several time instants, is shown in Fig. 3. Since the heat transfer coefficient for subcooled boiling is function of the wall temperature, it is observed that it will rise steadily until saturation is reached. After saturation, its value remains constant. When CHF occurs near the end of the channel, the heat removal degrades and the heat transfer coefficient suffers a severe drop as observed. This heat transfer crisis tends to travel to lower heights as time increases, which corresponds to a drying front propagation. For t = 4.0 s, the heat transfer coefficient drop is as large as 97%. When the power is cut off, the heat transfer coefficient drops along the rod due to the reduction of the rod superficial temperature.

Figure 4 shows the flow quality, where three different boiling regions can be observed, separated by the inflections of the curves. The first one separates the region of forced convection and subcooled boiling from the region of saturated boiling. The second inflection divides this last region from the post-dryout region. Near the entrance, a sharp increase in flow quality is observed, followed by a smooth increase when the CHF phenomenon occurs.

There is a reduction in the quality growth, indicating the position and instant of time where it takes place. This behavior can be explained by the reduction of the superficial flux and consequently a smaller increase of the enthalpy.

The maximum flow quality in this transient is equal to 0.57 at the outlet of the channel at t = 4.0 s.

The effect of the heat transfer regimes reflects on the clad temperature as shown in Fig.5. A maximum increase of 22% is seen in the clad temperatures when the boiling crisis phenomenon occurs. Clad melting can be avoided if the electric power supply is interrupted. Some clad points show temperature rise of 118 oC/s. As a result their temperatures could reach values as high as 1100oC in less than 7.0 s.

The propagation of the drying and rewetting fronts is represented in Figure 6, for several mass flow rates. For mass flow rate equal to 0.0535 m/s, the drying front has a mean velocity of 4.6 cm/s. For the rewetting one, this velocity is 2.4 cm/s. The front velocities presented in Figs.6, 7 e 8 are mean velocities which are calculated dividing the maximum distance reached by the front by the correspondent time interval. The mass flow influence on the velocities is also shown. Variations of + 20% and - 20% on the reference case ( = 0.0535 kg/s) are applied. It is noted that, as this variable rises, the drying velocity also rises from 4.6 to 4.9 cm/s, while the rewetting one goes from 2.4 to 4.4 cm/s. The CHF does not occur immediately after the power step, due to the radial and axial thermal resistances and condutances of the indirect heated rod. The time delay observed is reduced from 1.0 s to 0.8 s as the mass flow rate increases.

In order to observe the pressure influence on the reported velocities, the same transient is then analyzed under the coolant pressure of 8.0 MPa. Its inlet temperature is now equal to 280 oC. A 41% reduction in the pressure value has a strong influence on the rewetting front velocity: an increase of 936%. Otherwise there is a little influence on the drying one: it goes from 3.8 to 3.2 cm/s. These comparisons are shown in Fig. 7. The pressure also has considerable influence on the time delay which varies from 1.9 s to 1.4 s as the pressure changes from 8.0 MPa to 13.5 MPa.

The influence of the inlet mass flow rate was also investigated for the 8.0 MPa coolant pressure. Other authors have shown that this parameter has accentuated influence on the rewetting velocity at low pressures until 6.9 MPa. The test with a lower pressure was done to validate the model. The result obtained with the present model shown in Fig. 8 and confirms the trend. Note that the inlet mass flow rate has a smaller effect in the drying front. For a 20% reduction in mass flow the rewetting front velocity is reduced in 29.5%: It varies from 24.7 to 17.4 cm/s.

Conclusions

The present work analyzes the front propagation velocity for the drying out and rewetting processes, during the occurrence of critical heat flux in electrically heated simulators of nuclear fuel rods, caused by a power step.

This study is very important in the simulation of nuclear power plants as well as in metallurgical problems. At the occurrence of CHF, the amount of time required to cut off the electric power used in the heating of the simulator needs to be quantified. After the p ower cut off, the surface is rewetted when the temperature of the wall is less than the critical one. The two phenomena were analyzed individually by several authors (Carlson (1989),

Olek et al. (1988), Griffith et al.(1988)) and there were no information about the amount of time available for the operation of the protection systems.

Specially, there were few informations about rewetting, studied before only for descending films at low pressures. The work here presented supplies part of this lack of information.

Due to the radial and axial thermal resistances and capacitances of the indirect heating of the rod, the critical heat flux does not occur immediately after the power step. A certain time delay is observed. This time delay is reduced by increasing the pressure or the mass flow rate. In the beginning of the occurrence of CHF, the velocity of the propagation of the drying front is very high, being reduced gradually to an approximately constant value, around 4.6 cm/s. After the power cut off, the rewetting front presents a very large velocity of propagation, which is greatly affected by the system pressure and mass flow rate. The rewetting velocities were 24.7 cm/s and 2.4 cm/s for pressures of 8.0 MPa and 13.5 MPa, respectively. For the case of pressure of 8.0 MPa, the rewetting front propagation velocities were 17.4 cm/s and 24.7 cm/s for flow rates of 0.0471 kg/s and 0.0539 kg/s, respectively.

At the spot where the occurrence of CHF first starts, it was observed that the temperature increases at a rate of 118° C/s, which indicates that the wall temperature would reach its temperature limit, estimated around 1100° C, in approximately 7s. This is the amount of time available to turn off the electrical power supply. This observed heating rate is much larger than the value obtain by Mosaad (1988).

Additional details of the present work can be obtained in Lima (1997).

Acknowledgements The authors thank to FACEPE ( Fundaço de Apoio Cincia e Tecnologia do Estado de Pernambuco) for the support given to this work.

Appendix A

a) Forced convection: DITTUS - BOELTER s correlation (BJORNARD (1977)).

where k is the coolant thermal conductivity; De , the hydraulic diameter of the channel; Re, the Reynolds number; and Pr , the Prandtl one.

b) Nucleate boiling: THOM s correlation (TONG & WEISMAN, (1979)).

where Twall = external temperature of the cladding; Tsat = saturation temperature; q"SUP = heat flux; e p = pressure.

The transition between forced convection and nucleate boiling may be abrupt. This problem can be solved using the suggestion of Rohsenow(1961) that considers the heat flux divided into two parts:

where the first term refers to the convection in the absence of bubbles and the second is the heat transfer only affected by the bubble movement, without convection. In the present work the first term is calculated using href obtained from Eq. A.1 and the last one by Eq. A.2.

c) Critical heat flux: EPRI correlation (EPRI Report, 1983).

where A and C are constants which are dependent from pressure, is the local flow quality local, and , the inlet flow quality; is the local heat flux and q"CHF , the critical heat flux.

d) Transition boiling: Bjornard s correlation (BJORNARD,1977).

where q"TB is the heat flux in the transition boiling region, and q"MSFB is the heat flux at the Leidenfrost temperature (TMSFB ).

e) Minimum heat flux (Leidenfrost point): BJORNARD s correlation (1977).

where THN is the homogeneous nucleation temperature. This is the temperature at which the nucleation occurs spontaneously in the liquid in the absence of preferred nucleation sites. It is function of the pressure and can be predicted using standard nucleation theory. It can be obtained by the following expression of the TRAC -PF1 handbook:

where P = 3203.6 - P, ( in psia) and THN in 0F.

f) Film boiling: modified Groeneveld s correlation (BJORNARD, (1977)).

where: Y = two-phase flow factor of Miropol skiy, x = flow quality; Prwall =

Prandtl number evaluated at temperature Twall; G = coolant mass flux; De =

hydraulic diameter of the channel; g = dynamic viscosity - gaseous phase; g = coolant density - gaseous phase; f = coolant density - liquid phase; kg.= thermal conductivity -gaseous phase.

References BJORNARD, T. A.; GRIFFITH, P. PWR blowdown heat transfer. In:

Symposium on the thermal and hydraulics aspects of nuclear reactor safety, vol.1: Light Waters Reactors, pp. 17-39, 1977. [ Links ]

CARLSON, R.W., "Spreading of critical heat flux region during testing for onset of critical heat flux", Ann. Nucl. Energy, vol. 6, no. 2, pp. 49-62, 1989.

[ Links ]

GRIFFITH, P., MOHAMED, J. A. & BROWN, D., "Dryout front modeling for rod bundles", Nucl. Engin. and Design, vol. 105, pp. 223-229, 1988. [ Links ]

GUNNERSON, F. S. & YACKLE, T. R., "Quenching and rewetting of nuclear fuel rods", Nuclear Technology, vol. 54, pp. 113-117, 1981. [ Links ]

LIMA, R. DE C. F. DE, "Comportamento de vareta aquecida eletricamente durante transitório de fluxo critico de calor", Doctoral Thesis, Instituto de Pesquisas Energéticas e Nucleares / USP , So Paulo, 1997. [ Links ]

MOSAAD, M. Subcooled boiling heat transfer to flowing water in a vertical tube. Doctoral Thesis, Technischen Universitaet Berlin , 1988. [ Links ]

OLEK, S., ZVIRIN, Y. & ELIAS, E., "Rewetting of rod surfaces by falling liquid film as a conjugate heat transfer problem", Int. J. Multiphase Flow, vol. 14, no. 1, pp. 13-33, 1988. [ Links ]

PARAMETRIC STUDY of CHF data, volume 2. A generalized subchannel CHF correlation for PWR and BWR fuel assemblies. Prepared for Electric Power Research Institut, California, 1983. [ Links ]

SILVA NETO, A. J. da, ROBERTY, N.C., CARMO, E.G.D. CRISTE - um subcódigo para o cálculo da distribuiço axial, transiente, de temperaturas no canal de um reator PWR. Internal Report PEN-132, COPPE/UFRJ, Rio de Janeiro, 1983. [ Links ]

TONG, L.S. & WEISMAN, J. Thermal analysis of pressurized water reactors, American Nuclear Society, 1979. [ Links ]

TRAC-PF1. An advanced best-estimate computer program for pressurized water-reactor analysis. Safety Code Development Group Energy Division. [

Links ]

YU, S.K.W., FARMER, P. R. & CONEY, M.W.E., "Methods and correlations for the prediction of quenching rates on hot surfaces", Int. J. Multiphase Flow, vol. 3, pp. 415-448, 1977. [ Links ]

WBNP-4 4.4.2.2.4 Surface Heat Transfer Coefficients The fuel rod surface heat transfer coefficients during subcooled forced convection and nucleate boiling are presented in Section 4.4.2.8.1.

4.4.2.2.5 Fuel Clad Temperatures The outer surface of the fuel rod at the hot spot operates at a temperature of approximately 660°F for steady state operation at rated power throughout core life due to the onset of nucleate boiling.

Initially (beginning-of-life), this temperature is that of the clad metal outer surface.

During operation over the life of the core, the buildup of oxides and crud on the fuel rod surface causes the clad surface temperature to increase. Allowance is made in the fuel center melt evaluation for this temperature rise. Since the thermal-hydraulic design basis limits DNB, adequate-heat transfer is provided between the fuel clad and the reactor coolant so that the core thermal output is not limited by considerations of clad temperature.

4.4.2.2.6 Treatment of Peaking Factors The total heat flux hot channel factor, FQ, is defined by the ratio of the maximum to core average heat flux and is presented in Table 4.3-2 and discussed in Section 4.3.2.2.6.

This results in a peak local power of 5.52 kW/ft x FQ at full-power conditions. As described in Section 4.3.2.2.6, the peak linear power for determination of protection setpoints is 22.4 kW/ft.

The center line temperature at this kW/ft must be below the UO2 melt temperature over the lifetime of the rod, including allowances for uncertainties. The fuel temperature design basis is discussed in Subsection 4.4.1.2 and results in a maximum allowable calculated centerline temperature of 4700 °F. The peak linear power for prevention of centerline melt is > 22.4 kW/ft.

The centerline temperature at the peak linear power resulting from overpower transients/overpower errors (assuming a maximum overpower of 121%) is below that required to produce melting. Fuel centerline temperature at rated (100%) power and at the peak linear power for the determination of protection setpoints are presented in Table 4.4-1.

4.4.2.3 Critical Heat Flux Ratio or Departure from Nucleate Boiling Ratio and Mixing Technology The minimum DNBRs for the rated power, design overpower and anticipated transient conditions are given in Table 4.4-1. The minimum DNBR in the limiting flow channel will be downstream of the peak heat flux location (hot spot) due to the increased down stream enthalpy rise.

DNBRs are calculated by using the correlation and definitions described in the following Sections 4.4.2.3.1 and 4.4.2.3.2. The VIPRE-01 computer code (discussed in Section 4.4.3.4.1) is used to determine the flow distribution in the core and the local conditions in the hot channel for use in the DNB correlation. The use of hot channel factors is discussed in Section 4.4.3.2.1 (nuclear hot channel factors) and in Section 4.4.2.3.4 (engineering hot channel factors).

4.4-7

WBN Nuclear Power Generation/ATWS 1-FR-S.1 Unit 1 Rev. 0001 Step Action/Expected Response Response Not Obtained

19. CHECK Incore T/Cs less than IF Incore T/Cs are greater than 1200°F 1200°F. AND rising, THEN If this transition is missed, ** GO TO 1-SACRG-1, Severe then the following step will Accident Control Room Guideline Initial Response.

be used.

20. CHECK reactor subcritical: CONTINUE to borate.
a. Power range channels less than IF boration is NOT available, 5%. THEN ALLOW RCS to heat up to insert
b. Intermediate range startup rate negative reactivity from temperature NEGATIVE.

coefficients.

If one misses the IF red OR orange condition exists on transition to 1-SACRG-1, other Status Trees, then one will use 1-FR- THEN C.1. PERFORM actions of other FR Procedures which do not cool down or otherwise add positive reactivity to the core.

    • GO TO Step 4.
21. TERMINATE emergency boration:
a. PLACE BA transfer pumps in SLOW speed.
b. CLOSE emergency borate valve 1-FCV-62-138.
c. IF alternate boration opened, THEN Locally CLOSE 1-ISV-62-929.

Page 10 of 16

WBN Status Trees FR-0 Unit 1 Rev. 0014 If the transition is Attachment 1 missed at step 19, (Page 2 of 8) this first decisionMonitoring Critical Safety Functions block will indicate a red path. CORE COOLING FR-C Page 5 of 11

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]

Some examples of SRO exam items for this topic include:

  • Administrative processes for the installation of temporary instrumentation.
  • Processes for changing the plant or plant procedures.

Section IV provides an example of a satisfactory SRO-only question related to this topic.

D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

[10 CFR 55.43(b)(4)]

Some examples of SRO exam items for this topic include:

  • Process for gaseous/liquid release approvals, i.e., release permits.
  • Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.

Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.

SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.

E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.

The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:

Page 6 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

  • Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
  • Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
  • Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.

SRO-only knowledge should not be claimed for questions that can be answered solely using systems knowledge; e.g.:

  • how the system works.
  • system flow path.
  • component locations, etc.

SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:

  • the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
  • any AOP entry condition.
  • plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
  • immediate operator actions of a procedure.

Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.

Page 7 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, Yes RO question flowpath, logic, component location?

No Can the question be answered solely by knowing immediate operator actions? Yes RO question No Can the question be answered solely by knowing entry conditions for AOPs or plant parameters Yes RO question that require direct entry to major EOPs?

No Can the question be answered solely by knowing the purpose, overall sequence of events, or Yes RO question overall mitigative strategy of a procedure?

No Does the question require one or more of the following?

  • Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
  • Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps Yes SRO-only
  • Knowledge of diagnostic steps and decision points in the question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
  • Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

92.

Given the following conditions:

- Unit 1 is at 100% power.

- Maintenance personnel commence Appendix J, Penetration X-80 LLRT of 1-SI-30-701.

- In accordance with Appendix J, 1-FCV-30-37 and 1-FCV-30-40 are SHUT.

- Containment pressure approaches T/S LCO 3.6.4 limits.

Flow diagram of X-80 Which ONE of the following describes an action that will maintain containment pressure within the T/S LCO 3.6.4 limits?

____(1)____ AND in accordance with the ODCM, would be authorized by a _____(2)_____ release permit.

NOTE: 1-SI-30-701, Containment Isolation Valve Local Leak Rate Test Purge Air 1-FCV-30-37, LWR CNTMT PURGE EXH PRESS RLF 1-FCV-30-40, LWR CNTMT PURGE EXH PRESS RLF SOI-30.02, Containment Purge System SOI-65.02, Emergency Gas Treatment System T/S LCO 3.6.4, Containment Pressure A. (1) EGTS would be started in accordance with SOI-65.02 (2) weekly periodic B. (1) EGTS would be started in accordance with SOI-65.02 (2) conditional C. (1) Containment purge would be started in accordance with SOI-30.02 (2) weekly periodic D. (1) Containment purge would be started in accordance with SOI-30.02 (2) conditional

CORRECT ANSWER: D DISTRACTOR ANALYSIS:

A. Incorrect: As mentioned in the (D) distractor it is incorrect and plausible that the EGTS system be used for containment pressure control. Of note is the fact that the permitting requirements for an EGTS subsystem operation are component to 1-ODI-90-26.

B. Incorrect: While the need to obtain a prior to release or conditional permit is correct: the utilization of the EGTS system is not correct. As previously discussed, EGTS would assist in controlling containment pressure ONLY if the two dampers 1-FCV-30-37 and 1-FCV-30-40 were OPEN. A manual start of the EGTS subsystem (with no containment isolation phase A signal present) would assist in maintaining containment pressure because during such scenario, the containment and the annulus volumes would be cross-connected.

Therefore, choosing such distractor is plausible but incorrect as the containment pressure relief valves were closed in support of the LLRT.

C. Incorrect: Again, it is correct that the containment purge would be utilized for containment pressure control. It is not correct however, that the weekly permit resultant from 1-ODI-90-26 would authorize this discharge. It is plausible to believe this as the purge does discharge to the Units shield building stack and as 1-ODI-90-26 samples the shield building exhaust the implication is present that such procedure would enable the containment purge to be accomplished.

D. Correct: As seen in system description WBN-SDD-N3-30RB-4002, Reactor Building Ventilation System: The containment venting, for continuous pressure relief, is performed during modes 1-5, by opening the containment isolation (CI) valves FCV-30-40 and -37. This allows continuous venting of containment air into the Annulus through one of the Containment Vent Air Cleanup Units. This is also seen on flow print 1-47W866-1. Therefore, if the two FCVs mentioned are shut, then the pressure inside of containment will uncontrollably rise. As seen in system description N3-65-4001, Emergency Gas Treatment System: the EGTS establishes and keeps the annulus at a negative pressure and captures containment out-leakage. The EGTS is placed into service after a containment isolation (phase A) is received.

Furthermore, the flowpath through the EGTS subsystem shows that air is drawn from the annulus, processed through filter banks and then discharged with some flow being returned to the annulus and the remainder being sent out of the Units shield building exhaust stack.

This division of discharge maintains the annulus at a slight vacuum (relative to the isolated containment vessel). One may see that the EGTS subsystem is not designed to maintain the containment pressure within any bounds; it simply addresses any leakage which emanates from the isolated containment. Therefore, it is correct that initiating a containment purge will be the sole viable option for maintaining containment pressure within the limits of the Technical

Specifications.

As seen in Table 2.2-2 of the Offsite Dose Calculation Manual (ODCM), a containment purge requires that both the minimum sampling and analysis frequencies are P each purge. As seen in table 3.1 of the ODCM, this annotation indicates that a sample and analysis of the containment atmosphere must be completed prior to each release. As seen on print 1-47W866-1 (or on the simplified ICS screen shot), a containment purge is discharged to the Units shield building stack. Aforementioned was the fact that the EGTS subsystem also discharges (at least for a portion of its flow) to the shield building stack. Such a shield building stack discharge sampled and analyzed on a W frequency. This is contained again in table 2.2-2. A W frequency is at least once per 7 days. At WBN, the requirements of the ODCM are in part implemented by the Offsite Dose Instructions (ODIs). Of import to this question are two ODIs. 1-ODI-90-15, Containment Purge Release satisfies the requirements applicable to the containment purge in table 2.2-2. 1-ODI-90-26, Weekly Sampling Of Unit 1 Shield Building Exhaust addresses the ODCM compliance of the other effluents discharging through the Units shield building stack. The outputs of these ODIs are the release permits. Therefore, it is correct that before containment purge is to be initiated, that a release permit be authorized.

Question Number: 92 Tier: 2 Group: 2 K/A: 029 Containment Purge System A2 Ability to (a) predict the impacts of the following malfunctions or operations on the Containment Purge System; and (b) based on those predictions, use procedures to Correct: control, or mitigate the consequences of those malfunctions or operations:

A2.01 Maintenance or other activity taking place inside containment Importance Rating: 2.9 3.6 10 CFR Part 55: (CFR: 41.5 / 43.5 / 45.3 / 45.13) 10CFR55.43.b: 10 CFR 55.43(b)(2) and 10 CFR 55.43(b)(5)

K/A Match: K/A is matched because the applicant is required to predict the impact that the performance of Appendix J of 1-SI-30-701 (a maintenance activity which is conducted inside of the containment) has upon the containment purge system.

Namely, the applicant must identify that it would require that containment purge be started if containment pressure control became needed. Subsequently, the applicant must use the information in the ODCM to correctly select the required permitting to conduct the purge.

Technical

Reference:

Offsite Dose Calculation Manual (ODCM) 1-ODI-90-15, Containment Purge Release 1-ODI-90-26, Weekly Sampling Of Unit 1 Shield Building Exhaust 1-SI-30-701, Cntmt Isol Vlv Local LR Test Purge Air 1-47W866-1 ICS screenshot of EFF1 screen.

Proposed references to None be provided:

Learning Objective: 3-OT-SYS065A

9. Given plant conditions, IDENTIFY the applicable EGTS System limits and precautions related to the following:
b. SOI-65.02 Emergency Gas Treatment System
12. DESCRIBE the following aspects of TS and TRs
b. The Limiting Conditions for Operation, Applicability, and Bases.

Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question for the 2015-301 NRC SRO Exam Comments:

WBN System REACTOR BUILDING VENTILATION WBN-SDD-N3-30RB-4002 Description SYSTEM Rev. 0024 Document Page 50 of 96 3.1.3 Containment Air Return System (continued)

Ductwork associated with the fans consists of hydrogen collectors from the reactor cavity, the containment dome, shared collection headers from the lower compartment, the pressurizer compartment, and the steam generator compartments.

3.1.4 Containment Vent System The containment venting, for continuous pressure relief, is performed during modes 1-5, by opening the containment isolation (CI) valves FCV-30-40 and -37. This allows continuous venting of containment air into the Annulus through one of the Containment Vent Air Cleanup Units (CVACU)s, which are equipped with HEPA and charcoal filters. The airflow from containment into the Annulus is provided by the motive force of the differential pressure between the containment and the Annulus. This air mixes with the Annulus atmosphere before the AVC fan discharges it into the AB exhaust stack via the suction-side duct of the AB FHA exhaust fans. As an alternate to using the normal vent pathway, for containment pressure relief, either the pair of lower compartment purge lines (one supply and one exhaust), or one of the two pairs of upper compartment purge lines (one supply and one exhaust) may be used. The use of these alternate lines may require re-balancing of the supply duct airflow, as needed, to preclude a containment pressure rise. When an upper, or the lower, compartment purge line is used, the Containment Vent System must be isolated.

The Containment Vent System shall be isolated, during mode 6, by closing the valves FCV-30-40 and FCV-30-37 (Refer to subSection 4.20).

3.2 Component Description 3.2.1 Major Component Description Note: The following is vendor data which describe the performance characteristics for major system components. For more detailed information and component requirements, the appropriate contract should be referenced. The information included in this section shall be updated upon any modification, addition, or replacement of existing equipment. The data represent the manufacturers' rated capacities and not to be construed as required design values. Refer to Section 3.1 and Table 9.6 for design values.

A. Purge Supply Fans TVA Contract No. - 76K35-83246-1 Manufacturer - H. K. Porter Company, Incorporated Capacity - 14,000 cfm at 9.5" Static Pressure Type - Belt-Driven Centrifugal Motor - 50 hp Seismic - Category I B. Purge Exhaust Fans TVA Contract No. - 76K35-83246-1 Manufacturer - H. K. Porter Company, Incorporated

NPG System EMERGENCY GAS TREATMENT SYSTEM N3-65-4001 Description Rev. 0010 Document Page 26 of 52 3.1.2 Air Cleanup Unit (ACU) Subsystem The ACU subsystem is an ESF with two independent, 100% capacity trains. Each train consists of an exhaust fan, a HEPA-charcoal filter assembly, isolation valves, associated dampers and ductwork, and instruments and controls. The ACU fans are located in the auxiliary building EGTS room adjacent to the Unit 2 shield building on El 757.0. The ACU fan design flow rate and the annulus negative pressure to be maintained are shown in Table 7. The ACU intake is centrally located within the annulus above the steel containment dome. The intakes and ducting used to bring the air to ACU subsystem are shared with the AVCS.

The ACU subsystem starts automatically when a CIA is received. It provides two capabilities needed during a LOCA. One of these is the capability to reduce out-leakage of radioactive material from the shield building to within the guideline limits of Ref. 7.5.1. This is accomplished by establishing and keeping the annulus at a negative pressure (Ref. 7.4.3). The second capability is to capture containment out-leakage and process it through a series of HEPA and charcoal filters before release to the atmosphere. The ACU housing contains the following components listed in order: a moisture separator, a relative humidity heater, a prefilter, HEPA filter, two charcoal filter beds in series, and an after HEPA filter. An exhaust fan is provided downstream of each ACU housing. The air flow network can be aligned to exhaust annulus air through either EGTS filter train. This is accomplished by closing the AVCS isolation valves and opening the ACU subsystem valves. See Table 2 for a listing of valves and the valve alignment during the ACU operation. After the air cleanup subsystem has established the required annulus pressure, a maximum of 250 cfm of air is released through the shield building exhaust vent for a postulated single failure of one EGTS train or a maximum of 957 cfm for a postulated single failure of a control loop associated with one train of PCOs. (Ref 7.4.4). The remaining flow is recirculated in the annulus in a manner that promotes mixing, dilution, and holdup of the containment out-leakage. The recirculated air flow is discharged from a manifold extending completely around the bottom of the annulus. There are 23 ports in the manifold with a rated flow of 174 cfm each (Refs. 7.4.1 and 7.1.6). The vertical separation between the exhaust and the discharge ports is 168'-9".

After the air has been processed, the airflow network directs the air to redundant damper controlled flow dividers in the annulus. At this point, the flow network contains two airflow paths leading to the unit's shield building exhaust vent (either 1-PCO-65-80 and 1-PCV-65-81 or 1-PCO-65-82 and 1-PCV-65-83) and two airflow paths to the annulus manifold (either 1-PCO-65-88 and 1-PCV-65-86 or 1-PCO-65-89 and 1-PCV-65-87).

1-PCO-65-80, 82, 88 and 89 modulate to maintain the annulus pressure relative to the outside environment. The isolation dampers are zero leakage valves used to minimize outside air in-leakage from the shield building exhaust vent into the annulus. By varying the amount of air that is exhausted through the shield building exhaust vent, the negative annulus pressure is maintained. This pressure level is low enough so that leakage will be into the annulus from both primary containment and areas adjacent to the shield building.

The pressure differentials produced by wind effects and low temperature effects (Ref. 7.4.9) are also overcome by the appropriate selection of the pressure level.

The relative humidity heater and controls are arranged such that the heaters are energized whenever the EGTS ACU exhaust fans achieve a flow setpoint. The heaters de-energize when the fan flowrate is below the setpoint. Each heater is designed to maintain an air stream relative humidity of 70% before it is routed through the ACU filters in accordance with the requirements of Ref. 7.5.5.

NPG System EMERGENCY GAS TREATMENT SYSTEM N3-65-4001 Description Rev. 0010 Document Page 27 of 52 3.1.2 Air Cleanup Unit (ACU) Subsystem (continued)

Another feature incorporated into the ACU subsystem is the ability to cool the filters and adsorbers and remove radioactive decay heat in an inactive ACU containing radioactive material. This is accomplished with two crossover flow ducts that draw air at a minimum of 200 cfm through the active ACU from the discharge of the inactive ACU (Ref. 7.4.2). This flow rate is sufficient to limit the temperature rise in the inactive ACU to less than 75°F when even it is fully loaded (Ref. 7.4.2). Two butterfly valves are utilized in the crossover path to assure isolation. The isolation valves are opened automatically when the valve control switch is in P-AUTO position and one ACU fan is operating and the other ACU fan is idle.

These valves are normally closed and require operator action to be positioned in P-AUTO after an accident (see Section 4.2). However, the suction valve from the affected annulus to the inactive ACU must be opened by operator's action. Temperature rise is recorded in the MCR.

The two ACUs in the subsystem have steel housings. The housings incorporate a quench-type water spray and drain system for flooding the charcoal filters in case of fire.

(Ref. 7.2.23).

The EGTS must start within 30 seconds upon receiving a CIA signal (Refs. 7.4.4 and 7.4.22). The purge air valves (FCV-30-2, -5, -12, -54, -61, and -62) must close to meet this requirement (Ref. 7.4.22) 3.2 Component Description 3.2.1 Major Component Description EGTS related components are in Ref. 7.2.26 and 7.1.10.

The vendor data which describe basic design and performance characteristics for major system components are shown in Table 8. More detailed information and component requirements may be found in the contract drawing files. The data represents the manufacturer's rated capacities and should not to be construed as required design values.

3.2.2 Active Components Listing An active valves and dampers list is included in Table 3.

3.3 Instrumentation and Controls A detailed description of the EGTS electrical controls and logic can be found in Ref. 7.1.2.

and 7.1.3.

Operational limits, analytical limits, and safety limits, for instruments, as applicable, have been determined in Ref. 7.4.3. These limits have been used to establish instrument setpoints under all normal and LOCA conditions. The resulting setpoints are tabulated in the I-Tabs, 47B601-65 series drawings. (Ref. 7.1.8) 3.3.1 Instrumentation This section describes the instruments used to sense, indicate, and record flow, temperature, and pressure. Table 4 shows panel numbers for each instrument in both subsystems.

WBN OFFSITE DOSE CALCULATION MANUAL Revision 25 0 (ODCM) Page 27 of 195 Table 2.2-2-RADIOACTIVE GASEOUS WASTE MONITORING SAMPLING AND ANALYSIS PROGRAM*

(Page 1 of 3)

Gaseous Release Type Minimum Analysis Type of Activity Lower Limit of Sampling Frequency Analysis Detection Frequency (LLD) (Ci/ml)1 2

A. Waste Gas P P Noble Gases 1x10-4 Decay Tank Each Tank Each (Gamma Emitters)

Grab Sample Tank H-3 (oxide) 1x10-6 2

B. Containment P8 P Noble Gases 3

PURGE Each PURGE Each (Gamma Emitters) 1x10-4 Grab Sample Purge 2

C. Incore Instrument Each Each Noble Gases 1x10-4 3

Room PURGE PURGE9 Purge (Gamma Emitters)

Grab Sample D. Requirement Deleted 3,10 2 E. Auxiliary Building Exh. M M Noble Gases 1x10-4 F. Condenser Vacuum Exh.11 Grab Sample (Gamma Emitters)

G. Service Building Exh.

H. Deleted in Revision 11.

I. Deleted in Revision 11.

J. Deleted in Revision 11.

K. Auxiliary Building Exh. Continuous6 W H-3 (oxide) 1x10-6 L. Shield Building Exh. Tritium M. Condenser Vacuum Exh.11,12 Sample Continuous6 W7 I-131 1x10-12 Charcoal I-133 Sample 1x10-10 Continuous6 W7 Principal Gamma 1x10-11 Particulate Emitters2 Sample Continuous6 M Gross Alpha 1x10-11 Composite Particulate Sample Q Sr-89, Sr-90 1x10-11

  • See Table 3.1 (FREQUENCY NOTATION) for the surveillance frequency definitions.

WBN OFFSITE DOSE CALCULATION MANUAL Revision 24 0 (ODCM) Page 57 of 195 Table 3.1 - FREQUENCY NOTATION NOTATION FREQUENCY S At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

D At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

W At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 184 days.

3Q At least once per 276 days.

Y At least once per 365 days.

R At least once per 18 months.

N/A Not applicable.

P Completed prior to each release.

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]:

A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)]

Some examples of SRO exam items for this topic include:

  • Reporting requirements when the maximum licensed thermal power output is exceeded.
  • The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
  • National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
  • Processes for TS and FSAR changes.

Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic.

B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

Some examples of SRO exam items for this topic include:

  • Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
  • Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
  • Knowledge of TS bases that are required to analyze TS required actions and terminology.

SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.

SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. ROs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Page 3 of 16

The information Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010) required for the ODCM would be ACTIONS from the LCO and associated applicability statements tantamount to that (standardized TS; see example below) in the T/S which is "below the line."

RO knowledge Above this line Page 4 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)

(Tech Specs)

Can question be answered solely by knowing 1 Yes hour TS/TRM Action? RO question No Can question be answered solely by knowing the Yes LCO/TRM information listed above-the-line? RO question No Can question be answered solely by knowing the Yes TS Safety Limits? RO question No Does the question involve one or more of the following for TS, TRM, or ODCM?

  • Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
  • Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4) Yes SRO-only
  • Knowledge of TS bases that is required to analyze TS question required actions and terminology No The question uses Question might not be linked to "below the line" 10 CFR 55.43(b)(2) for SRO-only knowledge of the ODCM.

Page 5 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]

Some examples of SRO exam items for this topic include:

  • Administrative processes for the installation of temporary instrumentation.
  • Processes for changing the plant or plant procedures.

Section IV provides an example of a satisfactory SRO-only question related to this topic.

D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

[10 CFR 55.43(b)(4)]

Some examples of SRO exam items for this topic include:

  • Process for gaseous/liquid release approvals, i.e., release permits.
  • Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.

Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.

SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.

E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.

The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:

Page 6 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

  • Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
  • Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
  • Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.

SRO-only knowledge should not be claimed for questions that can be answered solely using systems knowledge; e.g.:

  • how the system works.
  • system flow path.
  • component locations, etc.

SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:

  • the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
  • any AOP entry condition.
  • plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
  • immediate operator actions of a procedure.

Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.

Page 7 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, Yes RO question flowpath, logic, component location?

No Can the question be answered solely by knowing immediate operator actions? Yes RO question No Can the question be answered solely by knowing entry conditions for AOPs or plant parameters Yes RO question that require direct entry to major EOPs?

No Can the question be answered solely by knowing the purpose, overall sequence of events, or Yes RO question overall mitigative strategy of a procedure?

No The question requires that the applicant select an Does the question require one or more of the following?

SOI section.

  • Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed
  • Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps Yes SRO-only
  • Knowledge of diagnostic steps and decision points in the question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
  • Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

93.

00:00:00 - Unit 1 is at 100% power.

- Unit 2 requests that SCCW be taken out of service to permit maintenance on the Unit 2 cooling tower (CT) basin.

0 1:00:00 - The crew performs the following step of Section 7.1, Shutdown, of SOI-27.03:

04:00:00 - Section 7.1 of SOI-27.03 is complete.

1 1 :00:00 - The Unit 2 CT basin is isolated and drained.

- Due to evaporative losses, the Unit 1 CT basin REQUIRES makeup.

Which ONE of the following describes which document for which the MINIMUM Flow is a basis AND the method for providing makeup to the Unit 1 CT?

The source of the MINIMUM flow requirement described in step [1] shown above is ____(1)____.

At 1 1 :00:0 1, the crew can provide makeup to the Unit 1 CT using ____(2)_____.

NOTE: SOI-27.03, Supplemental Condenser Circulating Water System 0-SOI-24.01, Raw Cooling Water System SOI-27.03, Section 8.7, Cooling Tower Basin Makeup with SCCW Shutdown 0-SOI-24.01 section 8.3.1, Unit 1 Bypass Strainer Operation-RCW Adjustment T/S LCO 3.7.9, Ultimate Heat Sink A. (1) the bases for T/S LCO 3.7.9 (2) ONLY the RCW system using section 8.3.1 of 0-SOI-24.01 B. (1) the NPDES permit for WBNP (2) ONLY the RCW system using section 8.3.1 of 0-SOI-24.01 C. (1) the bases for T/S LCO 3.7.9 (2) EITHER the SCCW supply using section 8.7 of SOI-27.03 OR the RCW system using section 8.3.1 of 0-SOI-24.01 D. (1) the NPDES permit for WBNP (2) EITHER the SCCW supply using section 8.7 of SOI-27.03 OR the RCW system using section 8.3.1 of 0-SOI-24.01

CORRECT ANSWER: B DISTRACTOR ANALYSIS:

A. Incorrect: Section 8.7 of SOI-27.03 uses the SCCW system to provide makeup to the cooling tower basins. It utilizes a valve named 0-FCV-27-112.

This valve admits water to the U2 cooling tower basin. Because the U2 cooling tower basin is drained for maintenance, it is correct that 0-FCV-27-112 could not be used to compensate for cooling tower evaporative loses; it is not correct and yet plausible to believe that the basis for T/S LCO 3.7.9 is the source for the step [1] cited from Section 7.1, Shutdown of SOI-27.03.

B. Correct: T/S LCO 3.7.9 stipulates that, The UHS shall be OPERABLE. The basis for this LCO indicates: The UHS is required to be OPERABLE and is considered OPERABLE if it contains water at or below the maximum temperature that would allow the ERCW System to operate for at least 30 days following the design basis LOCATo meet this condition, the UHS temperature should not exceed 85°F. The basis for this T/S does not mention any required river flow.

TVA must regulate discharges to the waters of the United States in accordance with the National Pollutant Discharge Elimination System (NPDES). As seen on page 8 of the TVA-Watts Bar Nuclear Plant NPDES Permit TN0020168, All changes to the flow rate of the SCCW discharge (Outfall 113) shall be done during periods when flow in the receiving waters is at a minimum of 3,500 cubic feet per second.The thermal mixing zone area has been modified and redefined for this permitThe discharge from Outfall 113 shall be limited and monitored by the permittee. Therefore, the source of the step [1] cited from Section 7.1, Shutdown of SOI-27.03 is the sites NPDES permit.

It is correct that ONLY the RCW system using section 8.3.1 of 0-SOI-24.01 can be used for cooling tower makeup.

C. Incorrect: While it is correct that the NPDES is the source for the step [1] cited from Section 7.1, Shutdown of SOI-27.03, it is not correct and yet plausible that 0-FCV-27-112 could be used to compensate for cooling tower evaporative loses.

D. Incorrect: It is correct that the NPDES permit is the source for the step [1] cited from Section 7.1, Shutdown of SOI-27.03.

As seen on print 1-47W-831-1, 0-FCV-27-112, admits water from upstream of the Watts Bar dam to the Unit 2 cooling tower basin. This water flows toward the Unit 2 CCW inlet via the Unit 2 flume. During makeup via this means the Unit 1 and Unit 2 flumes are cross connected. Therefore, the makeup admitted via 0-FCV-27-112 is provided to both Unit 2 and Unit 1. Therefore, it is incorrect that this mode of cooling tower makeup would be utilized (because the Unit 2 cooling tower is drained for maintenance). On the same print one may

observe the Unit 1 RCW discharge (30 line) to the Unit 1 cooling tower flume. This line not only provides the discharge of the RCW system but also any bypass strainer flow. The later was the cooling tower makeup afforded by the original plant design (e.g. before the installation of SCCW). The operating crew could divert some of the RCW supply directly to the cooling tower flumes to provide makeup to the cooling tower basins and thus maintain cooling tower levels. If one believed that 0-FCV-27-112 supplied the Unit 1 cooling tower basin, then this answer would be entirely correct.

Question Number: 93 Tier: 2 Group: 2 K/A: 075 Circulating Water System 2.1 Conduct of Operations 2.1.20 Ability to interpret and execute procedure steps.

Importance Rating: 4.6 4.6 10 CFR Part 55: (CFR: 41.10 / 43.5 / 45.12) 10CFR55.43.b: 10 CFR 55.43(b)(1)(2)

K/A Match: K/A is matched because the applicant must tell the meaning of (i.e.

interpret) a component of step [1] of Section 7.1, Shutdown of SOI-27.03. Additionally, the applicant must select the correct SOI section in order to provide makeup water to the cooling tower basin. Therefore, the applicant must interpret what the steps in the sections will perform.

Technical

Reference:

0-SOI-24.01, Raw Cooling Water System SOI-27.03, Supplemental Condenser Circulating Water System T/S LCO 3.7.9 T/S LCO 3.7.9 Basis 1-47W831-1 NPDES Permit for WBN Proposed references to None be provided:

Learning Objective: 3-OT-SYS027A

10. Given plant conditions, IDENTIFY the applicable Condenser Circulating Water System Precautions and Limitations related to the following:

SOI-27.01 Condenser Circulating Water System SOI-27.03 Supplemental CCW System Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question for the 2015-301 NRC SRO Exam Comments: See the marked up Clarification Guidance for SRO-only Questions.

UHS 3.7.9 Nothing above the 3.7 PLANT SYSTEMS line for this T/S leads one to the 3.7.9 Ultimate Heat Sink (UHS) correct answer.

LCO 3.7.9 The UHS shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. UHS inoperable. A.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AND A.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.9.1 Verify average water temperature of UHS is 85 F. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Watts Bar-Unit 1 3.7-21

UHS B 3.7.9 BASES APPLICABLE (Ref. 2), which requires a 30 day supply of cooling water in the UHS.

SAFETY ANALYSES (continued) The UHS satisfies Criterion 3 of the NRC Policy Statement.

LCO The UHS is required to be OPERABLE and is considered OPERABLE if it contains water at or below the maximum temperature that would allow the ERCW System to operate for at least 30 days following the design basis LOCA without the loss of net positive suction head (NPSH), and without exceeding the maximum design temperature of the equipment served by the ERCW System.

To meet this condition, the UHS temperature should not exceed 85 F.

APPLICABILITY In MODES 1, 2, 3, and 4, the UHS is required to support the OPERABILITY of the equipment serviced by the UHS and required to be OPERABLE in these MODES.

In MODE 5 or 6, the OPERABILITY requirements of the UHS are determined by the systems it supports.

ACTIONS A.1 If the UHS is inoperable, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.7.9.1 REQUIREMENTS This SR verifies that the ERCW System is available to cool the CCS to at least its maximum design temperature with the maximum accident or normal design heat loads for 30 days following a Design Basis Accident. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency (continued)

Watts Bar-Unit 1 B 3.7-49

TVA-Watts-Bar Nuclear Plant NPDES Permit TN0020168 Page 8 of 27 *

  • Discharges are authorized for Outfall 101 only during periods when flow in the receiving stream is at a minimum of 3,500 cubic feet per second. All changes to the flow rate of the SCCW discharge (Outfall 113) shall be done during periods when flow in the receiving waters is at a minimum of 3,500 Cubic feet per second. This includes periods of start-up, shutdown as well as other similar abrupt flow rate changes of the SCCW. When thermally loaded effluent is discharged through Outfall 102, all reasonable efforts shall be made to keep flow to a minimum of 3500 cubic feet per second in the receiving waters. If such flow is absent, the permittee shall verify protection of water quality by taking instream temperature measurements. Compliance with flow requirements for 3,500 cfs flow instream for Outfalls 101, 102 and 113 discharges shall be certified monthly with the submission of Discharge Monitoring Reports submitted to the Division for these outfalls. Records concerning the instream flow shall be maintained and available upon request.

The thermal mixing zone area has been modified and redefined for this permit; see diagram at Appendix 5H. The discharge from Outfall 113 shall be limited and monitored by the permittee as specified below:

  • In recognition of the dynamic behavior of the thermal effluent in the river, monitoring shall be required for an active mixing zone and a passive miXing zone as described in the permit rationale. The passive mixing zone includes the following dimensions:

(1) a maximum width of from bank to bank in the river, and (2) a maximum length of 1000 feet downstream of the outfall. It has been documented that there is a zone of (cool water) refuge in the bottom layer to allow for fish and other species to pass below the thermal plume. Compliance with the requirements below will be established for the active mixing zone at a maximum length of 2000 feet downstream of the outfall.

  • Compliance for the passive miXing zone shall be by two instream temperature surveys, one conducted during winter ambient conditions and one during summer ambient conditions. The surveys shall be performed while the SCCW system is thermally loaded with low river .flow conditions and shall include temperature profiles at a sufficient number of locations across the downstream edge of the passive mixing zone to locate the effluent plume. The measurements shall be compared with the results from the thermal plume model and shall be summarized in a report to the division semiannually.
  • Compliance with TEMPERATURE, Edge of Mixing zone; TEMPERATURE, Rise Upstream to Downstream; and TEMPERATURE, Rate of Change shall be applicable at the edge of the active mixing zone.
  • Daily maximum temperatures for the TEMPERATURE, effluent; TEMPERATURE, Edge of Mixing zone; TEMPERATURE, Rise Upstream to Downstream; and TEMPERATURE, Rate of Change shall be determined from 1-hour average values.

The average values shall be calculated every 15 minutes using the current and previous four 15-minute values, thus creating a rolling average.

  • As demonstrated by monitoring at the edge of the active mixing zone, the maximum temperature shall not exceed 30.5°C (except as a result of natural causes), the maximum change in temperature relative to the upstream control point shall not exceed 3°C (except as a result of natural causes), and the maximum temperature rate of change shall not exceed 2°C per hour (except as a result of natural causes).

asin Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]:

A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)]

Some examples of SRO exam items for this topic include:

  • Reporting requirements when the maximum licensed thermal power output is exceeded.
  • The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
  • National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
  • Processes for TS and FSAR changes.

Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic.

B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

Some examples of SRO exam items for this topic include:

The applicant must understand what

  • Application of Required Actions (Section 3) and Surveillance constitutes an Requirements (SR) (Section 4) in accordance with rules of application OPERABLE UHS. requirements (Section 1).
  • Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
  • Knowledge of TS bases that are required to analyze TS required actions and terminology.

SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.

SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. ROs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Page 3 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)

Nothing "above the line" for the UHS T/S leads the applicant to the RO correct answer.

knowledge Above this line Page 4 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)

(Tech Specs)

Can question be answered solely by knowing 1 Yes hour TS/TRM Action? RO question No Can question be answered solely by knowing the Yes LCO/TRM information listed above-the-line? RO question No Can question be answered solely by knowing the Yes TS Safety Limits? RO question No Does the question involve one or more of the following for TS, TRM, or ODCM?

  • Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
  • Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4) Yes SRO-only
  • Knowledge of TS bases that is required to analyze TS question required actions and terminology No Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only Page 5 of 16

94.

Given the following conditions:

- Unit 1 is stable in MODE 2.

- Chemistry reports that DOSE EQUIVALENT I-131 (DEI) is 1.0 µCi/gm.

Excerpt from T/S LCO 3.4.16 Which ONE of the following describes the application of the Technical Specifications for the conditions below?

In accordance with T/S LCO required action 3.4.16 A.1, DEI must be verified < ____(1)____ µCi/gm.

The NOTE LCO 3.0.4.c is applicable indicates that when an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made _____(2)_____.

(1) (2)

A. 14 when an allowance is stated in the individual value, parameter or other Specification B. 14 after performance of a risk assessment addressing inoperable systems and components, and establishment of risk management actions C. 21 when an allowance is stated in the individual value, parameter or other Specification D. 21 after performance of a risk assessment addressing inoperable systems and components, and establishment of risk management actions

CORRECT ANSWER: A DISTRACTOR ANALYSIS:

A. Correct: A verification that DEI is <14 µCi/gm must be made. Additionally, as discussed a risk assessment is not required to support the mode 1 change.

As seen in the required action A.1 for T/S LCO 3.4.16, a verification must be made that DEI is <14 µCi/gm. This limit is contained throughout the basis for T/S LCO 3.4.16.

LCO 3.0.4 states: When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made:

a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time;
b. After performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate; exceptions to this Specification are stated in the individual Specifications, or
c. When an allowance is stated in the individual value, parameter, or other Specification.

One may see that the required actions for condition A of T/S LCO 3.4.16 are modified by the NOTE which states: LCO 3.0.4.c is applicable. Because of this, entry into mode 1 is not impeded by the fact that condition A of T/S LCO is not met.

B. Incorrect: As seen in the required action A.1 for T/S LCO 3.4.16, a verification must be made that DEI is <14 µCi/gm. This limit is contained throughout the basis for T/S LCO 3.4.16. As such, the first half of this distractor is correct. It is plausible to believe that a risk assessment would be required because this would be the case if LCO 3.0.4 b were invoked to support the mode change.

C. Incorrect: It is not correct that a verification must be made that DEI is <21

µCi/gm. It is plausible to believe that this is the case because prior to amendment 91, T/S LCO 3.4.16 (and its basis) utilized this value. This amendment was placed into effect during the SRO applicants time in initial license training.

It is Correct: however, that a risk assessment is not required prior to an entry into mode 1.

D. Incorrect: It is not correct that a verification must be made that DEI is <21

µCi/gm. It is plausible to believe that this is the case because prior to amendment 91, T/S LCO 3.4.16 (and its basis) utilized this value. This amendment was placed into effect during the SRO applicants time in initial license training.

Additionally, as previously discussed, it is incorrect and yet plausible that a risk assessment be required to enter mode 1.

Question Number: 94 Tier: 3 Group:

K/A: 2.1 Conduct of Operations 2.1.34 Knowledge of primary and secondary plant chemistry limits.

Importance Rating: 2.7 3.5 10 CFR Part 55: (CFR: 41.10 / 43.5 / 45.12) 10CFR55.43.b: 10 CFR 55.43(b)(2)

K/A Match: K/A is matched because the applicant is required to possess the knowledge of one of the limits germane to RCS specific activity.

Technical

Reference:

T/S LCO 3.4.16 (current and a historical copy)

T/S LCO 3.4.16 basis T/S LCO 3.0.4 Proposed references to None be provided:

Learning Objective: 3-OT-TS-0304

1. Describe the LCO, Applicability and Bases for the LCO.

Cognitive Level:

Higher Lower X Question Source:

New X Modified Bank Bank Question History: New question for the 2015-301 NRC SRO Exam Comments:

LCO Applicability 3.0 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2.

LCO 3.0.2 Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6.

If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required unless otherwise stated.

LCO 3.0.3 When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to place the unit, as applicable, in:

a. MODE 3 within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />;
b. MODE 4 within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />; and
c. MODE 5 within 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />.

Exceptions to this Specification are stated in the individual Specifications.

Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by LCO 3.0.3 is not required.

LCO 3.0.3 is only applicable in MODES 1, 2, 3, and 4.

LCO 3.0.4 When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made:

a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time; (continued)

Watts Bar-Unit 1 3.0-1 Amendment 55

LCO Applicability 3.0 Notice that a,b and 3.0 LCO APPLICABILITY c or joined by OR.

LCO 3.0.4 b. After performance of a risk assessment addressing inoperable (continued) systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate; exceptions to this Specification are stated in the individual Specifications, or

c. When an allowance is stated in the individual value, parameter, or other Specification.

This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY.

LCO 3.0.6 When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, additional evaluations and limitations may be required in accordance with Specification 5.7.2.18, "Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

(continued)

Watts Bar-Unit 1 3.0-2 Amendment 55

RCS Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.16 The specific activity of the reactor coolant shall be within limits.

APPLICABILITY: MODES 1 and 2, MODE 3 with RCS average temperature (Tavg) 500°F.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. DOSE EQUIVALENT I-131 -----------------------NOTE--------------------

> 0.265 Ci/gm. LCO 3.0.4.c is applicable.

A.1 Verify DOSE EQUIVALENT I-131 Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 14 Ci/gm AND A.2 Restore DOSE EQUIVALENT 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> I-131 to within limit.

B. Gross specific activity of the B.1 Perform SR 3.4.16.2. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> reactor coolant not within limit.

AND B.2 Be in MODE 3 with 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Tavg < 500°F.

(continued)

Watts Bar-Unit 1 3.4-39 Amendment 41, 55, 91

RCS Specific Activity 3.4.16 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3 with 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time Tavg < 500°F.

of Condition A not met.

OR DOSE EQUIVALENT I-131

> 14 Ci/gm.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 Verify reactor coolant gross specific 7 days activity < 100/E Ci/gm.

SR 3.4.16.2 --------------------------------NOTE-----------------------------------

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT I-131 specific 14 days activity 0.265 Ci/gm.

AND Between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a THERMAL POWER change of 15% RTP within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period (continued)

Watts Bar-Unit 1 3.4-40 Amendment 41, 91

RCS Specific Activity 3.4.16 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.16.3 ----------------------------------NOTE---------------------------------

Required to be performed within 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Determine from a sample taken in MODE 1 after a 184 days minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Watts Bar-Unit 1 3.4-41

RCS Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.16 The specific activity of the reactor coolant shall be within limits.

APPLICABILITY: MODES 1 and 2, MODE 3 with RCS average temperature (Tavg) 500°F.

This is the OLD Limit ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. DOSE EQUIVALENT I-131 ---------------------NOTE--------------------

> 0.265 µCi/gm. LCO 3.0.4.c is applicable.

A.1 Verify DOSE EQUIVALENT I-131 Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 21 µCi/gm AND A.2 Restore DOSE EQUIVALENT 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> I-131 to within limit.

B. Gross specific activity of the B.1 Perform SR 3.4.16.2. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> reactor coolant not within limit.

AND B.2 Be in MODE 3 with Tavg < 500°F. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (continued)

Watts Bar-Unit 1 3.4-39 Amendment 41, 55

RCS Specific Activity 3.4.16 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3 with 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time Tavg < 500°F.

of Condition A not met.

OR DOSE EQUIVALENT I-131 >

21 µCi/gm.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 Verify reactor coolant gross specific 7 days activity < 100/E µCi/gm.

SR 3.4.16.2 --------------------------------NOTE-----------------------------------

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT I-131 specific 14 days activity 0.265 µCi/gm.

AND Between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a THERMAL POWER change of 15% RTP within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period (continued)

Watts Bar-Unit 1 3.4-40 Amendment 41

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]:

A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)]

Some examples of SRO exam items for this topic include:

  • Reporting requirements when the maximum licensed thermal power output is exceeded.
  • The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
  • National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
  • Processes for TS and FSAR changes.

Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic.

B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

Some examples of SRO exam items for this topic include:

  • Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
  • Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
  • Knowledge of TS bases that are required to analyze TS required actions and terminology.

SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.

SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. ROs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Page 3 of 16

Clarification Guidance for SRO-only Questions Question (in part) Rev 1 (03/11/2010) requires "below the ACTIONS from the LCO and associated applicability statements line" knowledge of (standardized TS; see example below) the T/S.

RO knowledge Above this line Page 4 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)

(Tech Specs)

Can question be answered solely by knowing 1 Yes hour TS/TRM Action? RO question No Can question be answered solely by knowing the Yes LCO/TRM information listed above-the-line? RO question No Can question be answered solely by knowing the Yes TS Safety Limits? RO question No Does the question involve one or more of the following for TS, TRM, or ODCM?

  • Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
  • Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4) Yes SRO-only
  • Knowledge of TS bases that is required to analyze TS question required actions and terminology No Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only Page 5 of 16

95.

In accordance with Tech Spec LCO 3.0.5, INOPERABLE equipment may be returned to service for the following reasons:

1. Demonstrate OPERABILITY of the equipment.
2. Demonstrate OPERABILITY of other Tech Spec required equipment.
3. Troubleshoot equipment to facilitate repair.

A. 1 ONLY B. 1 and 2 ONLY C. 1 and 3 ONLY D. 1, 2 and 3

CORRECT ANSWER: B DISTRACTOR ANALYSIS:

A. Incorrect: LCO 3.0.5 states: Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. Therefore, this answer is plausible because it is partly correct (i.e. that it is correct that LCO 3.0.5 may be used to demonstrate the OPERABILITY of the inoperable equipment) but it is not fully correct because the OPERABILITY of other equipment may be tested.

B. Correct: T/S LCO states:

Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment.

C. Incorrect: As mentioned, this distractor is incorrect. It is plausible because troubleshooting is part of the process which restores a failed component to an operable status. One may believe that the process by which a component is repaired and thus returned to an operable status is addressed by this T/S. This is not the case.

D. Incorrect: Again, while two of the three items listed in this distractor are correct, troubleshooting is not allowed by T/S LCO 3.0.5.

Question Number: 95 Tier: 3 Group:

K/A: 2.2 Equipment Control 2.2.21 Knowledge of pre- and post-maintenance operability requirements.

Importance Rating: 2.9 4.1 10 CFR Part 55: (CFR: 41.10 / 43.2) 10CFR55.43.b: 10 CFR 55.43(b)(2)

K/A Match: K/A is matched because the applicant is required to understand the use of T/S LCO 3.0.5 to an inoperable safety system or component.

Technical

Reference:

T/S LCO 3.0.5 Proposed references to None be provided:

Learning Objective: 3-OT-TS-0300

5. Given plant conditions where LCOs and/or TRs are not met, determine if equipment may be tested to demonstrate operability.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: Bank question taken verbatim from the last SQN NRC exam.

Comments:

LCO Applicability 3.0 3.0 LCO APPLICABILITY LCO 3.0.4 b. After performance of a risk assessment addressing inoperable (continued) systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate; exceptions to this Specification are stated in the individual Specifications, or

c. When an allowance is stated in the individual value, parameter, or other Specification.

This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY Note that of other equipment. This is an exception to LCO 3.0.2 for the system returned to troubleshooting is service under administrative control to perform the testing required to not allowed. demonstrate OPERABILITY.

Can Use this to retest:

1. inoperable equipment LCO 3.0.6 When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support OR system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, additional evaluations and limitations may be
2. other equipment.

required in accordance with Specification 5.7.2.18, "Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

(continued)

Watts Bar-Unit 1 3.0-2 Amendment 55

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

II. Some examples of additional knowledge and abilities as they pertain to an SRO license and the 10 CFR 55.43(b) topics [ES-401, Section D.1.c]:

A. Conditions and limitations in the facility license. [10 CFR 55.43(b)(1)]

Some examples of SRO exam items for this topic include:

  • Reporting requirements when the maximum licensed thermal power output is exceeded.
  • The required actions for not meeting administrative controls listed in Technical Specification (TS) Section 5 or 6, depending on the facility (e.g., shift staffing requirements).
  • National Pollutant Discharge Elimination System (NPDES) requirements, if applicable.
  • Processes for TS and FSAR changes.

Note: The analysis and selection of required actions for TS Sections 3 and 4 may be more appropriately listed in the following 10 CFR 55.43 topic.

B. Facility operating limitations in the TS and their bases. [10 CFR 55.43(b)(2)]

Some examples of SRO exam items for this topic include:

  • Application of Required Actions (Section 3) and Surveillance Requirements (SR) (Section 4) in accordance with rules of application requirements (Section 1).
  • Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 4.0.1 thru 4.0.4).
  • Knowledge of TS bases that are required to analyze TS required actions and terminology.

SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on knowledge of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action statements and the safety limits since Reactor Operators (ROs) are typically required to know these items.

SRO-only knowledge generally cannot be claimed for questions that can be answered solely based on expected RO TS knowledge. ROs are typically expected to know the LCO statements and associated applicability information, i.e., the information above the double line separating the Page 3 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

ACTIONS from the LCO and associated applicability statements (standardized TS; see example below)

RO knowledge Above this line Page 4 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

Figure 1: Screening for SRO-only linked to 10 CFR 55.43(b)(2)

(Tech Specs)

Can question be answered solely by knowing 1 Yes hour TS/TRM Action? RO question No Can question be answered solely by knowing the Yes LCO/TRM information listed above-the-line? RO question No Can question be answered solely by knowing the Yes TS Safety Limits? RO question No Does the question involve one or more of the following for TS, TRM, or ODCM?

  • Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)
  • Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4) Yes SRO-only
  • Knowledge of TS bases that is required to analyze TS question required actions and terminology No Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only Page 5 of 16

96.

In accordance with NPG-SPP-01.2, Administration of Site Technical Procedures, which ONE of the following describes the requirements for making a MINOR/EDITORIAL change to a technical procedure?

An Independent Qualified Review (IQR) ____(1)____ REQUIRED.

A 50.59 Screening Review ____(2)____REQUIRED.

A. (1) is (2) is B. (1) is (2) is NOT C. (1) is NOT (2) is D. (1) is NOT (2) is NOT

CORRECT ANSWER: B DISTRACTOR ANALYSIS:

A. Incorrect: NPG-SPP-01.2, Administration of Site Technical Procedures, 3.2.11.A describes that an AOR is not required but an IQR is and a 50.59 screening is not required. 50.59 screening requirements are directed by step 3.2.9.J B. Correct: NPG-SPP-01.2, Administration of Site Technical Procedures, 3.2.11.A describes how minor editorial changes do require an IQR to be performed but a 50.59 screening is not required.

C. Incorrect: NPG-SPP-01.2, Administration of Site Technical Procedures, 3.2.11.A describes that an AOR is not required but an IQR is 50.59 screening requirements are directed by step 3.2.9.J D. Incorrect: Administration of Site Technical Procedures, 3.2.11.A describes how minor editorial changes do require an IQR to be performed a 50.59 screening is not required. 50.59 screening requirements are directed by step 3.2.9.J

Question Number: 96 Tier: 3 Group:

K/A: G2.2 Equipment Control 2.2.6 Knowledge of the process for making changes to procedures Importance Rating: 3.0 3.6 10 CFR Part 55:

10CFR55.43.b: 10 CFR 55.43(b)(3)

K/A Match: K/A is matched because the applicant is required to demonstrate the knowledge of which reviews are required for a procedure change by the processes of the facility.

Technical

Reference:

NPG-SPP-01.2, Administration of Site Technical Procedures Proposed references to None be provided:

Learning Objective: 3-OT-AdminWB, NPG-SPP-01.02 Workbook 1-10 Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: Bank question G 2.2.6 97, last used on the 06/2011 NRC Exam.

Comments:

NPG Standard Administration of Site Technical NPG-SPP-01.2 Programs and Procedures Rev. 0011 Processes Page 16 of 54 3.2.9 Procedure Review Requirements (continued)

B. Procedures changing a QC holdpoint require an AOR by Quality Assurance.

C. A review for incorporation of NQAP requirements must be performed by Quality Assurance personnel or others knowledgeable of the QA requirements. This review will typically be performed as part of the Independent Qualified Review (IQR).

1. Quality Related procedures require technical adequacy review by an Independent Qualified Review (IQR). Form TVA 40667 (NPG-SPP-01.2-3, Procedure Verification Review Checklist) shall be used by the IQR for the review. IQR reviewers shall not be the person who prepared the procedure.
2. The responsible department manager shall select individuals to participate in the IQR program and these individuals shall complete site IQR training. During the procedure review process, the IQR reviewer shall identify any additional cross disciplinary review required to the procedure writer. See Section 3.2.24 for the qualification requirements for IQR.
3. When extensive changes that have been made (for example, 50% or more of the procedures) and technical change of content, a full-scope IQR is required.

D. PORC Review Required (per Technical Specification/NQAP). PORC review is required for the following:

1. New procedures or changes to existing procedures recommended by Regulatory Guide 1.33, Revision 2, Appendix A, February 1978; that require an evaluation in accordance with 10 CFR 50.59.
2. The emergency operating procedures which implement NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic Letter 82-33.
3. Physical Security Plan.
4. Radiological Emergency Plan.
5. Offsite Dose Calculation Manual.
6. Process Control Program (radwaste packaging and shipping).
7. Additional PORC reviews specifically required by site specific technical specifications or the plants licensing basis.
8. Proposed changes to TS; Technical Requirements Manual; their bases; amendments to the Operating License.
9. Selected 10 CFR 50.59 evaluations.
10. Selected 10 CFR 72.48 evaluations.

NPG Standard Administration of Site Technical NPG-SPP-01.2 Programs and Procedures Rev. 0011 Processes Page 18 of 54 3.2.9 Procedure Review Requirements (continued)

I. New technical procedures or changes to technical procedures that create or revise TEMPORARY MODIFICATIONS to structures, systems, and components (SSCs) for the purpose of utilizing the SSC for plant operation or crediting the SSC for performing a plant function shall be evaluated in accordance with NPG-SPP-09.5, Temporary Modifications, and shall be reviewed in accordance with NPG-SPP-09.4, 10 CFR 50.59 Evaluations of Changes, Tests, and Experiments and Step 3.2.9J and step 3.2.9L, as applicable. This does not include alterations for the purpose of testing the affected SSC.

J. New technical procedures and changes to technical procedures shall be reviewed to determine if the procedure is within the scope of 10 CFR 50.59 using Attachment 1 of NPG-SPP-09.4. The results of this determination shall be noted on the PCF (Form TVA 40665 - NPG-SPP-01.2-1) or BSL/ECM audit trail. This review shall be performed by a 10 CFR 50.59 qualified individual in accordance with NPG-SPP-09.4, 10 CFR 50.59 Evaluations of Changes, Tests, and Experiments. Procedures shall be evaluated for 10 CFR 50.59 applicability if those procedures contain information described in the UFSAR such as how structures, systems, and components are operated and controlled, including assumed operator actions and response times.

1. If it is determined that 10 CFR 50.59 is applicable to the procedure or the change being made, then a 10 CFR 50.59 screening review shall be performed in accordance with NPG-SPP-09.4 using Forms TVA 40518 (NPG-SPP-09.4-1, Applicability Determination/Screening Review/50.59 Evaluation Coversheet) and TVA 40673 (NPG-SPP-09.4-2, Screening Review Form). Per NPG-SPP-09.4, Form TVA 40517 (NPG-SPP-09.4-7, Procedure Change Evaluation) may be used if appropriate.
2. If, as the result of the 10 CFR 50.59 screening review, a 10 CFR 50.59 evaluation needs to be generated, ensure the evaluation is performed by a qualified reviewer in accordance with NPG-SPP-09.4, 10 CFR 50.59 Evaluations of Changes, Tests, and Experiments.

K. Changes to the Physical Security Contingency Plan, Radiological Emergency Plan and Implementing Procedures, and the NQAP do not require 10 CFR 50.59 screening reviews. Changes to these documents are made in accordance with 10 CFR 50.54, Conditions of Licenses.

NOTE Minor/editorial changes do not require 10 CFR 72.48 reviews. The 10 CFR 72.48 documents will be archived in EDM as stand alone documents.

L. New technical procedures or changes to technical procedures associated with Independent Spent Fuel Storage Installation (ISFSI) or shared (interfacing) systems which may impact ISFSI shall be reviewed to determine if the procedure or the change is within the scope of 10 CFR 72.48 in accordance with NPG-SPP-09.9, 10 CFR 72.48 Evaluations of Changes, Tests, and Experiments for Independent Spent Fuel Storage Installation.

NPG Standard Administration of Site Technical NPG-SPP-01.2 Programs and Procedures Rev. 0011 Processes Page 21 of 54 3.2.11 Minor/Editorial Changes A. Minor changes, such as inconsequential editorial corrections that do not affect the outcome, results, functions, processes, responsibilities, and requirements of the performance of procedure or instructions, require review by an IQR for quality-related procedures and approval by the appropriate approval authority. Minor changes do not require an AOR, 10 CFR 50.59 review, 10 CFR 72.48 review, or PORC review. Minor changes shall not change the intent of the procedure or alter the technical content or sequence of procedural steps.

B. Procedure changes that meet any of the following criteria are considered minor changes:

1. Correction of punctuation, style changes
2. Redundant or insignificant word or title changes
3. Correction of typographical errors including capitalization
4. Annotation of critical steps
5. Correction of reference errors
6. Omitted symbols that do not alter results
7. Incorrect units of measure due to editorial error
8. Misplaced decimals that are neither setpoint values nor tolerances
9. Page number discrepancies
10. Missing sign-offs, signatures, or date lines
11. Corrections to attachment identifiers
12. Corrections to titles of plant organizations, position titles, department/section/unit names when there is no change in authority, responsibility, or reporting relationships
13. Corrections to addresses, telephone numbers, or computer application names
14. Corrections to or additions of equipment nomenclature or locations in procedures to be consistent with approved drawings, documents, labels, or procedure content
15. Addition of or changes to equipment unique identifier information (unid) in procedures consistent with design output documents and which do not alter what component is operated
16. Corrections to or clarification of a note or precaution which does not alter the method of accomplishing a task

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]

Some examples of SRO exam items for this topic include:

  • Administrative processes for the installation of temporary instrumentation.
  • Processes for changing the plant or plant procedures.

Section IV provides an example of a satisfactory SRO-only question related to this topic.

D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

[10 CFR 55.43(b)(4)]

Some examples of SRO exam items for this topic include:

  • Process for gaseous/liquid release approvals, i.e., release permits.
  • Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.

Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.

SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.

E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.

The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:

Page 6 of 16

97.

Which ONE of the following describes the Containment Access authorization in accordance with TI-12.07A, Containment Access Modes 1 - 4?

General access into either the containment or the annulus; may be authorized provided that the incore flux detectors are in their normal storage location inside the crane wall

_____(1)_____ AND tagged with a _____(2)_____.

A. (1) ONLY (2) Hold Order B. (1) ONLY (2) Caution Order C. (1) OR approximately ten feet below the bottom of the core limit in any core thimble (2) Hold Order D. (1) OR approximately ten feet below the bottom of the core limit in any core thimble (2) Caution Order

CORRECT ANSWER: C DISTRACTOR ANALYSIS:

A. Incorrect: As seen in TI-12.07A, The incore flux detectors are in either of the two approved storage locations, as described in TI-41, (Normal STORAGE in Crane Wall or approximately ten feet below bottom of core limit in any core thimble). A Hold Order must be issued to the RADIATION PROTECTION Shift Supervisor and in place on the incore detector drive motors to prevent operation while personnel are inside Containment or Annulus.

Note that a hold order uses Danger Tags to indicate the isolation points of the hold order.

It is plausible to believe this distractor is correct because one may recollect the one approved storage location but not the other.

B. Incorrect: It is incorrect and yet plausible that the incores only have one approved storage location. It is also incorrect that a caution tag would be used to secure such. It is plausible to believe this as in the vast majority of the references to tagging the incores, TI-12.07A is mute as to the type of tag used. It simply states that the incores are to be TAGGED.

C. Correct: There are two approved storage locations for the incores. Also, as aforementioned, a Danger Tag is used.

D. Incorrect: While it is correct that there are two approved storage locations for the incores, it is not correct and yet plausible that a caution tag would be used to secure the incore detectors.

Question Number: 97 Tier: 3 Group:

K/A: 2.3 Radiation Control 2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Importance Rating: 3.2 3.7 10 CFR Part 55: (CFR: 41.12 / 45.9 / 45.10) 10CFR55.43.b: 10 CFR 55.43(b)(4)

K/A Match: K/A is matched because the applicant is required to demonstrate one of the SRO duties applicable to containment entry.

Technical

Reference:

TI-12.07A, Containment Access Modes 1-4 Proposed references to None be provided:

Learning Objective: 3-OT-TI-1207, Containment Access

12. Discuss the precaution associated specifically to an entry into the annulus and lower containment.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: Bank question G 2.3.12 97 used on the 09/2010 Sequoyah NRC exam.

Comments:

WBN Containment Access TI-12.07A Unit 1 Modes 1 - 4 Rev. 0007 Page 17 of 50 3.2.3 Operations A. The Shift Manager (SM) or designee (SRO) reviews and authorizes all General Access entries into Containment or Annulus by completion of Appendix A Section 1.0, Authorization for General Access, after ensuring the following:

1. The incore flux detectors are in either of the two approved storage locations, as described in TI-41, (Normal STORAGE in Crane Wall or approximately ten feet below bottom of core limit in any core thimble). A Hold Order must be issued to the RADIATION PROTECTION Shift A hold order is a Supervisor and in place on the incore detector drive motors to prevent danger tag. operation while personnel are inside Containment or Annulus.[C5,6]
2. Access Control Custodian established OR airlock door alarms are enabled.
a. The requirement for establishing an Access Control Custodian may be waived by the SM in the event that urgent entry is required. In these cases the airlock door alarm will remain enabled.
3. The work activity or evolution is approved to be performed, including authorization of entries inside the Polar Crane Wall when below Mode 2.

B. IF personnel require access inside the Polar Crane Wall while in Mode 1 or 2 OR require entry when the incore detectors are NOT TAGGED or are NOT properly stored, the SM (concurrent with the RP Manager) must evaluate the necessity of the entry, issue special instructions (if any), and authorize the entry by completion of Appendix A, Section 2.0 Authorization for Special Access.

Such entries require issuance of a special ALARA Plan and approvals in accordance RCI-128, ALARA Program. [C5,6]

C. WHEN the radiological hazard associated with a Special Access entry is no longer present, THEN the Shift Manager (concurrent with the RP Manager) may relax the entry requirements allowing the return to General Access by the completion of Appendix A, Section 3.0, Exit From Special Access Requirements.

D. The Shift Manager ensures that in the event of an evacuation from Upper Containment through Lower Containment that the incore detectors are placed in a safe condition (tagged or stored) prior to authorizing personnel to open the Personnel Hatch #2 (Subhatch, 757) for exit from Containment.

E. The Access Control Custodian will be briefed on responsibilities and expectations for the implementation of this TI. The briefing as a minimum is to cover items contained in Step 3.2.1. C and D.

WBN Containment Access TI-12.07A Unit 1 Modes 1 - 4 Rev. 0007 Page 7 of 50 2.2 Developmental References (continued)

I. RCI-128, ALARA Program Implementation J. TI-134, Control of Portable Two Way Radios K. TI-229, Temporary Shielding Program L. WBN FSAR Questions 22.26, 212.116, and 212.129 3.0 PRECAUTIONS AND LIMITATIONS 3.1 General Precautions and Limitations A. All access portals (Airlocks el 757/716, equipment hatches el 757, and Annulus el 713) to the Containment building SHALL be controlled to prevent unauthorized entry while in Modes 1 through 4. Radiation Protection (RP) shall maintain positive access control of Containment and Annulus in accordance with RP procedures.

B. All entries into Containment or Annulus while in Modes 1 through 4 are authorized by completion of applicable sections of Appendix A, Containment/Annulus Entry Authorization.

C. One Appendix A is required for each area entered (Upper Containment, Lower Containment, Annulus). Appendix As are typically issued for a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period only. However the SM may approve an extension beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Authorization requirements for entries are described as follows:

1. General Access is authorized by the completion of Appendix A, Section 1.0, which requires the incore detectors to be TAGGED and in either of the two approved storage locations, as described in TI-41, (Normal STORAGE in Crane Wall or approximately ten feet below bottom of core limit in any core thimble). This ensures that a detector is in a location which would not expose entry personnel. In addition, this section also permits entry inside the Polar Crane Wall when below Mode 2.
2. Special Access is authorized by completion of Appendix A, Section 2.0, for any entry into Containment or Annulus when the incore detectors are NOT TAGGED or NOT in their approved storage location. Special Access authorization is also required for entries inside the Polar Crane Wall or Regen Hx Room during Modes 1 or 2.
3. All entries requiring Special Access Authorization shall require approvals in accordance with RCI-128, ALARA Program Implementation to ensure that appropriate controls are established to prevent personnel overexposure (i.e., Special RWP and APR).

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]

Some examples of SRO exam items for this topic include:

  • Administrative processes for the installation of temporary instrumentation.
  • Processes for changing the plant or plant procedures.

Section IV provides an example of a satisfactory SRO-only question related to this topic.

D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

[10 CFR 55.43(b)(4)]

Some examples of SRO exam items for this topic include:

The requirement

  • Process for gaseous/liquid release approvals, i.e., release permits.

tested in this

  • Analysis and interpretation of radiation and activity readings as they question is one in pertain to selection of administrative, normal, abnormal, and emergency which ONLY a SRO procedures.

can sign for. Also, Analysis and interpretation of coolant activity, including comparison to the SRO is the one emergency plan criteria and/or regulatory limits.

who will sign for (in eSOMS) the holder SRO-only knowledge should not be claimed for questions that can be of the clearance. answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.

E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.

The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:

Page 6 of 16

98.

Given the following conditions:

- Unit 1 is at 100% power.

- A Containment entry INSIDE of the Polar Crane wall is REQUIRED.

Which ONE of the following describes the Containment entry in accordance with TI-12.07A, Containment Access Modes 1-4?

____(1)____ can approve the containment entry listed above.

The personnel hazard stipulated by TI-12.07A can be mitigated by the crew reducing

____(2)____.

NOTE: 1-GO-4, Normal Power Operation section 5.3 of 1-GO-4, 5.3 Unit Shutdown from 100% to 30% Reactor Power SOI-30.02, Containment Purge System A. (1) ONLY the SM (2) reactor power in accordance with 1-GO-4 B. (1) ONLY the SM (2) containment temperature in accordance with SOI-30.02 C. (1) EITHER the SM or the US (2) reactor power in accordance with 1-GO-4 D. (1) EITHER the SM or the US (2) containment temperature in accordance with SOI-30.02

CORRECT ANSWER: A DISTRACTOR ANALYSIS:

A. Correct: As seen in section 3.2.3 of TI-12.07A, it is true that the Shift Manager (SM) or designee (SRO) reviews and authorizes all General Access entries into ContainmentIF personnel require access inside the Polar Crane Wall while in Mode 1 or 2the SM (concurrent with the RP Manager)authorize the entry by completion of Appendix A, Section 2.0 Authorization for Special Access. Note that Appendix A (page 2 of 3) does not allow a designee to sign for the shift manager (thus authorizing the special access to containment).

TI-12.07A lists that All entries requiring Special Access Authorization shall require approvalsto ensure that appropriate controls are established to prevent personnel overexposure. Therefore, the concern at hand with respect to TI-12.07A is radiological dose.

Radiological dose can be reduced by reducing reactor power.

Therefore it is correct that the concern explicated in TI-12.07A may be mitigated by reducing power using 1-GO-4.

B. Incorrect: As described, the SM provides the approval for the containment entry.

It is not correct that the crew would purge the containment to mitigate the concern of TI-12.07A as this TI indicates that the concern at hand is personnel overexposure.

By practical experience, lower containment is hot. By the Unit T/S, lower containment is over 100°F (when in Mode 1). Because of this, one must be concerned with heat stress and stay time calculations.

Inevitably what would be done to correct this would be to run containment purge. However, this would be done to satisfy the concerns of the safety manual procedure, TVA-TSP-18.906, Heat Stress. TI-12.07A does not contain any reference to the temperature of containment.

It is plausible to believe that a purge would be conducted because such would be the case; again this would be done to satisfy the heat stress requirements found in the mentioned chapter of the safety manual and not those of TI-12.07A. Also amplifying this plausibility is the fact that TI-12.07B, Containment Access Modes 5&6 does regard heat stress as a concern and ensures that persons conducting closeout inspections of the containment (for entry into Mode 4) are briefed on heat stress.

C. Incorrect: While it is correct that a power reduction would mitigate the stated concern of TI-12.07A for the containment entry, it is not correct that either the SM or the US could authorize the entry. As discussed, only the SM can approve a special containment entry. It is plausible to believe this as the US can sign for a general containment access.

D. Incorrect: As previously discussed, it is incorrect and yet plausible that a

containment purge would mitigate the concern of TI-12.07A. Also, it is incorrect and yet plausible that either the SM or the US could authorize the entry.

Question Number: 98 Tier: 3 Group:

K/A: 2.3 Radiation Control 2.3.14 Knowledge of radiation or contamination hazards that may arise during normal,l abnormal, or emergency conditions or activities.

Importance Rating: 3.4 3.8 10 CFR Part 55: (CFR: 41.12 / 43.4 / 45.10) 10CFR55.43.b: 10 CFR 55.43(b)(4) and 10 CFR 55.43(b)(5)

K/A Match: K/A is matched because the applicant is required to exhibit knowledge of the hazards present during a special containment entry and then compare such to those presented in TI-12.07A to arrive at a mitigation strategy.

Technical

Reference:

TI-12.07A, Containment Access Modes 1-4 TI-12.07B, Containment Access Modes 5-6 Proposed references to None be provided:

Learning Objective: 3-OT-TI-1207, Containment Access

12. Discuss the precaution associated specifically to an entry into the annulus and lower containment Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question for the 2015-301 NRC SRO Exam Comments:

WBN Containment Access TI-12.07A Unit 1 Modes 1 - 4 Rev. 0007 Page 7 of 50 2.2 Developmental References (continued)

I. RCI-128, ALARA Program Implementation J. TI-134, Control of Portable Two Way Radios K. TI-229, Temporary Shielding Program L. WBN FSAR Questions 22.26, 212.116, and 212.129 3.0 PRECAUTIONS AND LIMITATIONS 3.1 General Precautions and Limitations A. All access portals (Airlocks el 757/716, equipment hatches el 757, and Annulus el 713) to the Containment building SHALL be controlled to prevent unauthorized entry while in Modes 1 through 4. Radiation Protection (RP) shall maintain positive access control of Containment and Annulus in accordance with RP procedures.

B. All entries into Containment or Annulus while in Modes 1 through 4 are authorized by completion of applicable sections of Appendix A, Containment/Annulus Entry Authorization.

C. One Appendix A is required for each area entered (Upper Containment, Lower Containment, Annulus). Appendix As are typically issued for a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period only. However the SM may approve an extension beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Authorization requirements for entries are described as follows:

1. General Access is authorized by the completion of Appendix A, Section 1.0, which requires the incore detectors to be TAGGED and in either of the two approved storage locations, as described in TI-41, (Normal STORAGE in Crane Wall or approximately ten feet below bottom of core limit in any core thimble). This ensures that a detector is in a location which would not expose entry personnel. In addition, this section also permits entry inside the Polar Crane Wall when below Mode 2.
2. Special Access is authorized by completion of Appendix A, Section 2.0, for any entry into Containment or Annulus when the incore detectors are NOT This represents the TAGGED or NOT in their approved storage location. Special Access radiological hazard authorization is also required for entries inside the Polar Crane Wall or of the special Regen Hx Room during Modes 1 or 2.

access - personnel overexposure. 3. All entries requiring Special Access Authorization shall require approvals in accordance with RCI-128, ALARA Program Implementation to ensure that appropriate controls are established to prevent personnel overexposure (i.e., Special RWP and APR).

WBN Containment Access TI-12.07A Unit 1 Modes 1 - 4 Rev. 0007 Page 12 of 50 3.2.2 Personnel Entering/Exiting Containment A. Personnel needing to enter Containment or Annulus must request that RP initiate an Appendix A, to obtain authorization for entry. One Appendix A is required for each area entered (Upper Containment, Lower Containment, Annulus) and is typically issued for a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.

B. Personnel must notify RP Shift Supervision in advance of planned entries to coordinate support and establish access requirements.

C. General entries into Containment or Annulus on days other than designated Containment days, requires the approval of the Shift Manager or designee and RP Superintendent or designee. The Work Week Manager should also be consulted to confirm the need for entry AND may be contingent on the following:

  • Activity was scheduled prior to T-0 to be performed on requested day, OR
  • Activity is emergent and High Priority, i.e., LCO, WO priority 1 or 2, Ops concern, Management concern, etc.

AND

  • Activity cannot be rescheduled to work on another Containment entry day due to in-progress surveillance, NRC late date, FEG week, etc. AND
  • Resources are available to support entry and work.

D. The Access Control Custodian is responsible for ensuring completion and submittal of paperwork to Operations for closure.

E. All personnel must obtain an RWP briefing prior to entering Containment or Annulus, including a briefing on the alternate evacuation route through Personnel Hatch #2 (Subhatch, 757) between Upper and Lower Containment.

If the Upper Airlock is inoperable and an alternate evacuation route must be taken, RP and Operations must be contacted to ensure the incore detectors are tagged and properly stored prior to opening Personnel Hatch #2 (Subhatch, 757) between Upper and Lower Containment.

F. Upon initial entry into Containment or Annulus the airlock/access telephones are to be checked for proper operation and documented on Appendix B, Section 4.0, Airlock Phone Checks.

G. Entries into Containment or Annulus require that personnel be accounted for at all times. Accountability is maintained utilizing Appendix B, Personnel Accountability Logsheet(s). These logs are maintained in the following manner:

WBN Containment Access TI-12.07A Unit 1 Modes 1 - 4 Rev. 0007 Page 17 of 50 3.2.3 Operations A. The Shift Manager (SM) or designee (SRO) reviews and authorizes all General Access entries into Containment or Annulus by completion of Appendix A Section 1.0, Authorization for General Access, after ensuring the following:

1. The incore flux detectors are in either of the two approved storage locations, as described in TI-41, (Normal STORAGE in Crane Wall or approximately ten feet below bottom of core limit in any core thimble). A Hold Order must be issued to the RADIATION PROTECTION Shift Supervisor and in place on the incore detector drive motors to prevent operation while personnel are inside Containment or Annulus.[C5,6]
2. Access Control Custodian established OR airlock door alarms are enabled.
a. The requirement for establishing an Access Control Custodian may be waived by the SM in the event that urgent entry is required. In these cases the airlock door alarm will remain enabled.
3. The work activity or evolution is approved to be performed, including authorization of entries inside the Polar Crane Wall when below Mode 2.

B. IF personnel require access inside the Polar Crane Wall while in Mode 1 or 2 OR require entry when the incore detectors are NOT TAGGED or are NOT properly stored, the SM (concurrent with the RP Manager) must evaluate the necessity of the entry, issue special instructions (if any), and authorize the entry by completion of Appendix A, Section 2.0 Authorization for Special Access.

Such entries require issuance of a special ALARA Plan and approvals in accordance RCI-128, ALARA Program. [C5,6]

C. WHEN the radiological hazard associated with a Special Access entry is no longer present, THEN the Shift Manager (concurrent with the RP Manager) may relax the entry requirements allowing the return to General Access by the completion of Appendix A, Section 3.0, Exit From Special Access Requirements.

D. The Shift Manager ensures that in the event of an evacuation from Upper Containment through Lower Containment that the incore detectors are placed in a safe condition (tagged or stored) prior to authorizing personnel to open the Personnel Hatch #2 (Subhatch, 757) for exit from Containment.

E. The Access Control Custodian will be briefed on responsibilities and expectations for the implementation of this TI. The briefing as a minimum is to cover items contained in Step 3.2.1. C and D.

WBN Containment Access TI-12.07A Unit 1 Modes 1 - 4 Rev. 0007 Page 31 of 50 Appendix A (Page 1 of 3)

Containment/Annulus Entry Authorization 1.0 AUTHORIZATION FOR GENERAL ACCESS Personnel Access To Be Used:(3) 757 Airlock (Upper) 716 Airlock (Lower) 713 Annulus Airlock Door Alarms Enabled? YES Initials NO(4) Initials Custodian Designated? YES Initials NO(4)(6) Initials Incore Detectors Tagged and Stored IAW TI-41 and NO YES Initials entry inside the Polar Crane Wall while in Mode 2 or above.

Containment/Annulus Access Authorization: Mode(s): ____________

(1)

Entry Authorization:

Unit SRO or SM Date Time (1)

Entry Authorization:

RP Supt. or designee Date Time Remarks:

(1) RP Superintendent or designee and SM/SRO signature required for all Containment entries.

(2) Entry inside the Polar Crane Wall or Regen Hx Room while in Modes 1 & 2 REQUIRES authorization by RP Manager AND SM in accordance with RCI-128 prior to entry.

(3) IF Upper or Lower Containment entry required with incore probes NOT tagged or NOT Stored IAW TI-41, THEN Shift Manager AND RP Manager must authorize by completion of Section 2.0, Special Access Authorization.5, 6 Requirement for Upper Containment entry based on safety concern for Containment egress if upper airlock doors become inoperable.

(4) Signature verifies availability of Access Control Custodian for Containment, if Airlock door alarms are not enabled.

(5) Radiological hazard NO longer present - irradiated source NO longer poses potential for personnel exposure (Incore detectors TAGGED and properly stored OR removed) AND protective measures/controls are established.

(6) The SM may waive establishing Access Custodian in the event of urgent entry.

WBN Containment Access TI-12.07A Unit 1 Modes 1 - 4 Rev. 0007 Page 32 of 50 Appendix A (Page 2 of 3)

Containment/Annulus Entry Authorization 2.0 AUTHORIZATION FOR SPECIAL ACCESS Personnel Access To Be Used:(3) 757 Airlock (Upper) 716 Airlock (Lower) 713 Annulus Airlock Door Alarms Enabled? YES Initials NO(4) Initials Custodian Designated(3) YES Initials Authorized Work Group(s): ___________________________________________________________________

Description of Authorized Work: _______________________________________________________________

Work Limitations or Special Instructions (i.e., Access Control Requirements, specific ALARA Plan, etc.):

Incore Detectors NOT Tagged or Polar Crane Wall (Modes 1 or 2)

NOT Stored IAW TI-41(3) Initials or Regen Hx Rm Entry Required(2) Initials Containment/Annulus Access Authorization: Mode(s): ______________

Entry Authorization - Incore Detectors NOT TAGGED, NOT Stored IAW TI-41 OR Crane Wall entry required while in Mode 1 or 2.(2)(3)

Shift Manager Date Time Entry Authorization - Incore Detectors NOT TAGGED, NOT Stored IAW TI-41 OR Crane Wall entry required while in Mode 1 or 2.(2)(3)

RP Manager Date Time (1) RP Superintendent or designee and SM/SRO signature required for all Containment entries.

(2) Entry inside the Polar Crane Wall or Regen Hx Room while in Modes 1 & 2 REQUIRES authorization by RP Manager AND SM in accordance with RCI-128 prior to entry.

(3) IF Upper or Lower Containment entry required with incore probes NOT tagged or NOT Stored IAW TI-41, THEN Shift Manager AND RP Manager must authorize by completion of Section 2.0, Special Access Authorization.5, 6 Requirement for upper containment entry based on safety concern for containment egress if upper airlock doors become inoperable.

(4) Signature verifies availability of Access Control Custodian for Containment, if Airlock door alarms are not enabled.

(5) Radiological hazard NO longer present - irradiated source NO longer poses potential for personnel exposure (Incore detectors TAGGED and properly stored OR removed) AND protective measures/controls are established.

(6) The SM may waive establishing Access Custodian in the event of urgent entry.

WBN Containment Access TI-12.07B Unit 0 Modes 5 & 6 Rev. 0006 Page 12 of 38 Date ________

4.1 Preliminary Actions (continued)

[8] ENSURE personnel who will be performing inspections are briefed on the following:

  • personnel safety precautions and protective measures (climbing, heat stress, ALARA, etc)
  • detailed acceptance criteria in Section 5.0 for housekeeping/cleanliness standards and expectations (Appendix E may be utilized in conducting inspections)
  • techniques for conducting an inspection, where to look and what to check, etc.
  • potential consequences of inadequate inspection on sump operability and impact to plant operation (i.e., leaking PZR Bypass Spray Valves)
  • requirements for documenting deficiencies which cannot be immediately corrected ________

SRO

[9] IF known glycol leak or RCP oil leak exists, THEN PERFORM 1-TI-12.20, Containment Formaldehyde Stay Time Calculation. ________

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]

Some examples of SRO exam items for this topic include:

  • Administrative processes for the installation of temporary instrumentation.
  • Processes for changing the plant or plant procedures.

Section IV provides an example of a satisfactory SRO-only question related to this topic.

D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

[10 CFR 55.43(b)(4)]

Some examples of SRO exam items for this topic include:

Again, this question

  • Process for gaseous/liquid release approvals, i.e., release permits.

is questioning the

  • Analysis and interpretation of radiation and activity readings as they requirements for pertain to selection of administrative, normal, abnormal, and emergency containment entry procedures.

(an SRO function). Analysis and interpretation of coolant activity, including comparison to The special access emergency plan criteria and/or regulatory limits.

into containment requires the SRO-only knowledge should not be claimed for questions that can be authorization of the answered solely based on RO knowledge of radiological safety principles; SM. e.g., RWP requirements, stay-time, DAC-hours, etc.

E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.

The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:

Page 6 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

  • Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
  • Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
  • Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.

SRO-only knowledge should not be claimed for questions that can be answered solely using systems knowledge; e.g.:

  • how the system works.
  • system flow path.
  • component locations, etc.

SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:

  • the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
  • any AOP entry condition.
  • plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
  • immediate operator actions of a procedure.

Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.

Page 7 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, Yes RO question flowpath, logic, component location?

No Can the question be answered solely by knowing immediate operator actions? Yes RO question No Can the question be answered solely by knowing entry conditions for AOPs or plant parameters Yes RO question that require direct entry to major EOPs?

No Can the question be answered solely by knowing the purpose, overall sequence of events, or Yes RO question overall mitigative strategy of a procedure?

No Using knowledge of the contents of Does the question require one or more of the following? TI-12.07A, the applicant must

  • Assessing plant conditions (normal, abnormal, or select between two emergency) and then selecting a procedure or section of a SOI sections.

procedure to mitigate, recover, or with which to proceed

  • Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps Yes SRO-only
  • Knowledge of diagnostic steps and decision points in the question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
  • Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

99.

Given the following timeline:

00:00:00 Unit 1 is at 100% power.

A FIRE occurs in the B trained SDBD rooms.

00:0 1:00 Unit 1 remains STABLE at 100% power.

The following indications develop:

NOTE: 1-AOI-30.1, Plant Fires 1-AOI-30.2, Fire Safe Shutdown Which ONE of the following describes the proper implementation of the AOIs and EOPs?

A. At 00:0 1:00, 1-AOI-30.1 would continue to be in effect, because ONLY ONE train of safety related equipment is at risk.

B. At 00:00:00, 1-AOI-30.2 would be entered and would take precedence over the EOP set, SOLELY because the fire occurred in the SDBD rooms.

C. At 00:0 1:00, 1-AOI-30.2 would be entered and would NOT take precedence over the EOP set, because the Appendix R fire does not analyze for subsequent casualties.

D. At 00:0 1:00, 1-AOI-30.2 would be entered and would take precedence over the EOP set, because plant indications demonstrate that the ability of the plant to achieve and maintain safe shutdown is jeopardized.

CORRECT ANSWER: D DISTRACTOR ANALYSIS:

A. Incorrect: At 00:00:00, 1-AOI-30.1 is in effect as there is a fire in the plant. At 00:0 1:00, conditions exist which indicate that the fire in the SDBD room has caused multiple spurious actuation (a spurious start of the 1B-B SIP and the opening of the B trained BIT valve). This indicates to the SRO that a loss of plant control is imminent and step 4 of the 1-AOI-30.1 will direct that 1-AOI-30.2 be entered to address the Appendix R fire. It is plausible to believe that 1-AOI-30.1 will continue to be in effect because one may claim that single failure criteria exists and that the loss of one train of any component does not cause a loss of safety function. One may also consider the drastic actions which are taken during an Appendix R fire and thus wait until more severe impacts are observed.

B. Incorrect: 1-AOI-30.2 does state that: For an Appendix R fire, this procedure [1-AOI-30.2] takes precedence over the Emergency Operating procedures. Therefore, this portion of the distractor is correct.

However, one must note that the spurious actuations occur at 00:0 1:00. As described, 1-AOI-30.1 does not immediately transition one to 1-AOI-30.2 if a fire develops in the SDBD room. 1-AOI-30.1 contains the guidance that the SRO must observe: 1. Multiple spurious actuations of systems/components, 2. Erratic or questionable indications on numerous MCR meters/recorders or 3. Multiple trains/channels of safety related equipment involved. Then the SRO must decide that a loss of plant control is imminent and transition to 1-AOI-30.2.

This distractor is incorrect because at the very onset of the fire (at time 00:00:00) it claims that 1-AOI-30.2 is in effect because of the location of the fire. While there is a section of 1-AOI-30.2 to address a fire in the SDBD rooms, it is only used after an entry is made into 1-AOI-30.2 from 1-AOI-30.1 after the criteria aforementioned are noted.

C. Incorrect: The background seen in 1-AOI-30.2 (pg 12) relates: I. No other accident is assumed to occur concurrently with a fire. Therefore, while it is true that the Appendix R fire does not analyze for subsequent casualties, as mentioned previously it is not true that the EOP set takes precedence over 1-AOI-30.2.

D. Correct: At 00:0 1:00, indications exist that support an entry into 1-AOI-30.2. As seen in step 3 of 1-AOI-30.1, the fire has demonstrated the potential to affect plant control (safe shutdown capability). As already mentioned, 1-AOI-30.2 takes precedence over the EOP set.

Question Number: 99 Tier: 3 Group:

K/A: G 2.4.23 2.4 Emergency Procedures / Plan Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations.

Importance Rating: 3.6 4.4 10 CFR Part 55: (CFR: 41.7 / 41.10 / 43.5 / 45.12) 10CFR55.43.b: 10 CFR 55.43(b)(5)

K/A Match: K/A is matched because the applicant is required to understand the basis behind the preferential implementation of the appropriate procedure set (either the emergency or abnormal operating procedures) for various casualties.

Technical

Reference:

1-AOI-30.1, Plant Fires 1-AOI-30.2, Fire Safe Shutdown Proposed references to None be provided:

Learning Objective: 3-OT-AOI-3000, Plant Fires

4. Given a set of plant conditions, DESCRIBE operator actions required in accordance with 1-AOI-30.

Cognitive Level:

Higher X Lower Question Source:

New X Modified Bank Bank Question History: New question for the 2015-301 NRC SRO Exam Comments: The question is SRO only as detailed in the distractor analysis. The question meets the general SRO only criteria of Assessment of facility conditions an selection of appropriate procedures during normal, abnormal, and emergency situations.

WBN Plant Fires 1-AOI-30.1 Unit 1 Rev. 0002 Step Action/Expected Response Response Not Obtained 3.0 OPERATOR ACTIONS

1. IF a valid verbal report, annunciation or indication of a fire is present, THEN REQUEST UO perform Appendix B.
2. ENSURE both trains Control Room REFER TO SOI-31.01, Control Building Isolation (CRI) signals are DARK on HVAC System, to evaluate aligning Master Status Panel (MSP) on M-6. CREVS suction to the other side of the Control Building.

NOTE The decision to trip the Unit and declare an Appendix R fire is left to the judgment of the Unit SRO/Shift Manager and must be based on the magnitude of the fire and its potential effect on the equipment/components necessary to achieve and maintain cold shutdown.

3. MONITOR magnitude of the fire and The concern is the fire's the potential to affect plant control ability to affect the (safe shutdown capability): capability of the plant to achieve and maintain a

systems/components.

  • Erratic or questionable indications on numerous MCR meters/recorders. These are criteria used by
  • Multiple trains/channels of safety the SRO to determine if an related equipment involved. Appendix R fire is to be declared (i.e. if a transition
4. IF loss of plant control is imminent or to 1-AOI-30.2 is warranted).

becomes imminent during the performance of this Instruction, THEN The stem of the

    • GO TO AOI-30.2, Fire Safe question shows Shutdown. that a loss of plant control is imminent.

Page 4 of 11

WBN Fire Safe Shutdown 1-AOI-30.2 Unit 1 Rev. 0004 Step Action/Expected Response Response Not Obtained 4.0 OPERATOR ACTIONS NOTES

  • The decision to trip the unit and declare an Appendix R fire is left to the judgment of the Unit SRO/SM and must be based on the 1-AOI-30.2 trumps magnitude of the fire and its potential effect on the equipment and the EOP set. components necessary to achieve and maintain cold shutdown
  • For an Appendix R fire, this procedure takes precedence over the Emergency Operating Procedures
  • AUO local operator actions should be assigned as early as possible by an SRO or UO NOT involved with immediate actions of this procedure.
1. DETERMINE the fire location has RETURN to 1-AOI 30.1 the potential to affect equipment needed for safe shutdown.

NOTE For a fire that touches a soft interface (NO physical wall or barrier), as indicated by heavy dashed lines in 1-AOI-30.2 APP B, choose the room where the fire is predominate. When the fire is basically centered between the rooms the actions of either room are sufficient.

2. REFER to 1-AOI-30.2 APP B, Notice that while the SDBD Elevation Diagrams, to determine rooms have a C-series applicable 1-AOI-30.2 C-Series appendix associated with them, appendix.

a fire within such rooms does not automatically cause the use of this procedure.

Page 4 of 22

WBN Fire Safe Shutdown 1-AOI-30.2 Unit 1 Rev. 0004 5.1 Background and Assumptions (continued)

G. The Safe Shutdown Logic Diagram also shows the paths available to provide the safety functions for the safe shutdown conditions described in Paragraphs 5.1E and 5.1F. For each safety function, the equipment required to accomplish the safety function has been divided into Keys which represent groups of functionally-related equipment necessary to accomplish the safety function. These are also represented on the Safe Shutdown Logic Diagram.

H. At least one path of equipment or components needed to achieve safe shutdown is required to remain operable or capable of being operated for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following a postulated fire (to establish long-term heat removal via RHR).

I. No other accident is assumed to occur concurrently with a fire therefore, a valid SI signal is assumed not to be present at the time of an Appendix R fire.

However, spurious SI signal actuation could occur as a result of the effects of the fire. Since many of the actions in the Safe Shutdown analysis require No other accident components to be in positions opposite that required by SI, a spurious SI would is assumed to require these components to be repositioned.

occur with the Appendix R fire. For example, the BIT outlet valves are required to be closed for an Appendix R fire. The purpose of this is to:

1. Guarantee flow to the RCP seal line for boron injection
2. Prevent pressurizer overfill (no RCS break is assumed and normal charging/letdown may not be available due to fire or loss of air).
3. Prevent damage to the charging pump due to fast drawdown of the VCT (automatic circuit for the swap over to RWST on low VCT level is not guaranteed).

J. In general it is assumed that shutdown of the plant will be performed from the Main Control Room for a postulated fire elsewhere in the plant. For shutdown from outside the Control Building, it is essential that, functionally, the same equipment and instrumentation be available from the Aux Control Room or remote stations or otherwise be justified. Loss of offsite power is assumed concurrently with MCR fires.

K. The possibility of shutdown and cooldown of the plant from the auxiliary control room was considered in the manual actions of the approved Site Engineering calculation.

L. Where the spurious operation of a component could defeat the required system safety function, manual actions are taken to address the effects of spurious component operation. Components identified as those which could prevent a safe shutdown should they spurious operate are those that:

Page 12 of 22

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]

Some examples of SRO exam items for this topic include:

  • Administrative processes for the installation of temporary instrumentation.
  • Processes for changing the plant or plant procedures.

Section IV provides an example of a satisfactory SRO-only question related to this topic.

D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

[10 CFR 55.43(b)(4)]

Some examples of SRO exam items for this topic include:

  • Process for gaseous/liquid release approvals, i.e., release permits.
  • Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.

Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.

SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.

E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.

The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:

Page 6 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

  • Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
  • Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
  • Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.

SRO-only knowledge should not be claimed for questions that can be answered solely using systems knowledge; e.g.:

  • how the system works.
  • system flow path.
  • component locations, etc.

SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:

  • the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
  • any AOP entry condition.
  • plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
  • immediate operator actions of a procedure.

Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.

Page 7 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, Yes RO question flowpath, logic, component location? Note that this question does not No question the entry condition for the 1-Can the question be answered solely by knowing immediate operator actions? Yes RO question (as such AOI-30.2 procedure is No entered by transitioning from 1-AOI-30.1).

Can the question be answered solely by knowing entry conditions for AOPs or plant parameters Yes RO question that require direct entry to major EOPs?

No Can the question be answered solely by knowing This question requires the purpose, overall sequence of events, or Yes RO question that the applicant overall mitigative strategy of a procedure?

understand the transition No from 1-AOI-30.1 to 1-AOI-30.2 AND the Does the question require one or more of the following? prioritization of the appendix R fire recovery

  • Assessing plant conditions (normal, abnormal, or with respect to the EOP emergency) and then selecting a procedure or section of a set.

procedure to mitigate, recover, or with which to proceed

  • Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps Yes SRO-only
  • Knowledge of diagnostic steps and decision points in the question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
  • Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16

100.

Which ONE of the following describes the Protective Action Recommendations (PARs)?

In accordance with EPIP-4, Site Area Emergency, PARs are ____(1)____.

In accordance with EPIP-1, Emergency Plan Classification Logic, ____(2)____

can assume the responsibility for PARs when the respective emergency center is staffed and operational.

A. (1) optional (2) CECC director B. (1) optional (2) TSC RP Manager C. (1) NOT made (2) CECC director D. (1) NOT made (2) TSC RP Manager

CORRECT ANSWER: C DISTRACTOR ANALYSIS:

A. Incorrect: IAW REP Generic and EPIP-5 GE; PARs are only made at the General Emergency Level. REP Generic and EPIP-4, SAE do not direct a PAR be made.

IAW EPIP-1 and CECC EPIP-1, the CECC Director is the ONLY person that may assume the responsibility for PARs.

B. Incorrect: IAW REP Generic and EPIP-5 GE; PARs are only made at the General Emergency Level. REP Generic and EPIP-4, SAE do not direct a PAR be made.

There are responsibilities that the CECC Director may delegate; this is not one of those responsibilities.

The TSC RP Manger makes recommendations to the SED and CECC Director on Protective requirements and dose management.

C. Correct: IAW REP Generic and EPIP-5 GE; PARs are only made at the General Emergency Level. REP Generic and EPIP-4, SAE do not direct a PAR be made.

There are responsibilities that the CECC Director may delegate; this is not one of those responsibilities.

D. Incorrect: IAW REP Generic and EPIP-5 GE; PARs are only made at the General Emergency Level. REP Generic and EPIP-4, SAE do not direct a PAR be made.

There are responsibilities that the CECC Director may delegate; this is not one of those responsibilities.

The TSC RP Manger makes recommendations to the SED and CECC Director on Protective requirements and dose management.

Question Number: 100 Tier: 3 Group:

K/A: 2.4 Emergency Procedures / Plan 2.4.44 Knowledge of emergency plan protective action recommendations.

Importance Rating: 2.4 4.4 10 CFR Part 55: (CFR: 41.10 / 41.12 / 43.5 / 45.11) 10CFR55.43.b: 10 CFR 55.43(b)(5)

K/A Match: KA is matched because the question requires knowledge of emergency plan protective action recommendations.

Technical

Reference:

REP Generic EPIP-5, GE EPIP-4, SAE EPIP-1, Emergency Plan Classification Logic CECC EPIP-1 Proposed references to None be provided:

Learning Objective: 3-OT-PCD-048C, Radiological Emergency Plan

8. Given a plant situation that requires PARs, determine the correct PARs IAW EPIP-5.

Cognitive Level:

Higher Lower X Question Source:

New Modified Bank Bank X Question History: Bank question 2.4.44 700, last used on the WBN 08/2010 NRC exam.

Comments:

Radiological RADIOLOGICAL EMERGENCY PLAN REP-Generic Emergency (GENERIC PART) Rev. 0104 Procedure Page 33 of 90 5.2.2 Alert (continued)

1. Class of emergency.
2. Type of actual or projected release (airborne, waterborne, or surface spill) and estimated duration/impact times.
3. Estimate of quantity of radioactive material released or being released and the height of release.
4. Chemical and physical form of released material, including estimates of the relative quantities and concentration of noble gases, iodines, and particulates.
5. Prevailing weather (wind velocity, direction, temperature, atmospheric stability data, and form of precipitation, if any).
6. Actual or projected doses at site boundary.
7. Projected dose rates and integrated dose at about 2, 5, and 10 miles, including sector(s) affected.
8. Estimate of any surface spill radioactive contamination.
9. Emergency response actions underway.
10. Request for any needed onsite support by offsite organizations.
11. Prognosis for worsening or termination of event based on plant information.

G. The JIC may be activated.

H. Periodic media releases are provided.

I. The SED augments plant shift personnel, as necessary, to initiate corrective and protective actions.

5.2.3 Site Area Emergency Upon declaration of this class:

A. All the actions performed in section 5.2.2 are performed.

B. Personnel knowledgeable of plant systems are dispatched to the SEOC. Upon notification, these individuals should arrive at the applicable emergency operations center within a timeframe limited only by their commuting time.

C. Any appropriate protective actions for the public are recommended to State agencies by the CECC.

D. The JIC is activated.

Radiological RADIOLOGICAL EMERGENCY PLAN REP-Generic Emergency (GENERIC PART) Rev. 0104 Procedure Page 34 of 90 5.2.4 General Emergency Upon declaration of this class:

A. All the actions performed in section 5.2.3 are performed.

B. Appropriate protective action recommendations to the State are required upon declaration of General Emergency.

C. If this is the initial classification, the MCR notifies the local government agencies within 15 minutes, and passes along the protective action recommendations.

5.3 Transportation Accidents 5.3.1 Notification by Carrier In the event of a transportation accident involving a TVA shipment of radioactive materials, the carrier (or other person at the accident site) contacts the ODS. The carrier has procedures outlining the notifications.

5.3.2 Notification of ODS A. State B. EDO C. Shift Manager of the Affected Site D. CECC Director E. Radiological Assessment Manager 5.3.3 CECC Director Actions The CECC Director notifies the NRC, DOT, State authorities, ANI, and DOE (information only). The appropriate State agency, NRC, ANI, and DOE have duty officers available 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day to facilitate notification of their respective agencies.

5.3.4 Radiological Assessment Manager Actions The Radiological Assessment Manager will dispatch a radiological monitoring team, if deemed necessary by the CECC Director or requested by the appropriate State agency. A Radioactive Material Specialist may be sent with the team. The TVA Representative at the scene will be the senior TVA person at the site of the incident.

WBN General Emergency EPIP-5 Unit 0 Rev. 0043 Page 12 of 34 3.7 Notification of the Nuclear Regulatory Commission (NRC)

CAUTION Notification of the NRC shall be completed immediately after notification of the appropriate State or local agencies and not later than one hour after the time of Emergency Classification.

[1] COMPLETE Appendix D, Notification of the NRC.

NOTE When the TSC is staffed, the open and continuous line of communications with the NRC may be transferred to the NRC Coordinator position.

[2] IF REQUESTED by the NRC, DIRECT a member of the Operations staff (SRO if Available) to maintain an open and continuous line of communications as directed by NRC.

3.8 Monitor / Re-evaluate the Event

[1] Monitoring and reevaluation of plant events along with communicating significant changes should be performed continuously as a function of the emergency response. Methods used to communicate significant changes are not formalized and may vary depending upon staffing levels as well as availability of personnel or equipment.

Appendix E provides a systematic approach to monitoring/re-evaluation and the communication of significant changes in plant conditions.

CONTINUE to conduct State follow-ups until the CECC has assumed State communications responsibilities using Appendix F General Emergency Follow-up Information Form to communicate follow-up information.

[2] Reevaluation of significant changes must additionally include a determination of whether Protective Action Recommendations (PARs) should be upgraded.

The need to upgrade PARs is determined through the continuous assessment of Appendix H Initial Protective Action Recommendations.

IF it has been determined that a PAR Upgrade is appropriate, THEN COMMUNICATE the Upgrade to the State using Appendix J Upgrade-Protective Action Recommendation.

[3] CONTINUE to assess PARs until the CECC has assumed PAR responsibilities.

WBN General Emergency EPIP-5 Unit 0 Rev. 0043 Page 15 of 34 Appendix A (Page 1 of 1)

General Emergency Initial Notification Form

1. This is a Drill This is an Actual Event - Repeat - This is an Actual Event
2. The SED at Watts Bar has declared a GENERAL EMERGENCY
3. EAL Designator:_______________________(Use only one EAL)
4. Radiological Conditions: (Check one under both Airborne and Liquid column.)

Airborne Releases Offsite Liquid Releases Offsite Minor releases within federally approved limits1 Minor releases within federally approved Releases above federally approved limits1 limits1 Release information not known Releases above federally approved limits1 (1Tech Specs/ODCM) Release information not known (1Tech Specs/ODCM)

5. Event Declared: Time:__________________ Date:_____________________
6. The Meteorological Conditions are: (Use 46 meter data from the Met Tower. IF data is NOT available from the MET tower, contact the National Weather Service by dialing 9-1-423-586-8400. The National Weather Service will provide wind direction and wind speed.)

Wind Direction is FROM:_____________degrees Wind Speed: ______________m.p.h (15 minute average) (15 minute average)

7. Provide Protective Action Recommendation utilizing Appendix H: (Check either 1 or 2 or 3)

Recommendation 1 WIND Recommendation 2 RECOMENDATION 1 RECOMENDATION 2 FROM DEGREES EVACUATE LISTED SECTORS (Mark wind EVACUATE LISTED SECTORS (2 mile Radius and 10 miles downwind) direction (2 mile Radius and 5 miles from step 6) downwind)

SHELTER remainder of 10 mile EPZ SHELTER remainder of 10 mile EPZ CONSIDER issuance of POTASSIUM CONSIDER issuance of IODIDE in accordance with the State Plan POTASSIUM IODIDE in accordance with the State Plan A1, B1, C1, D1, C7, C9, D2, D4, D5, D6, D7, D8, D9 From 26-68 A1, B1, C1, D1, C7, D2, D4, D5 A1, B1, C1, D1, A3, A4, D2, D3, D4, D5, D6, D7, D8, D9 From 69-110 A1, B1, C1, D1, A3, D2, D4, D5 A1, B1, C1, D1, A2, A3, A4, A5, A6, A7, D2, D3, D5, D6 From 111-170 A1, B1, C1, D1, A2, A3, D2, D5 A1, B1, C1, D1, A2, A3, A5, A6, A7, B2, B3, B4, B5, C2 From 171-230 A1, B1, C1, D1, A2, A3, B2, B4, C2 A1, B1, C1, D1, B2, B3, B4, B5, C2, C3, From 231-270 A1, B1, C1, D1, B2, B4, C2 A1, B1, C1, D1, B2, B3, C2, C3, C4, C5, C6, C11 From 271-325 A1, B1, C1, D1, B2, C2, C4, C5, A1, B1, C1, D1, C2, C4, C5, C6, C7, C8, C9, C10, C11, D4, D9 From 326-25 A1, B1, C1, D1, C2, C4, C5, C7, C8, D4 Recommendation 3 SHELTER all sectors CONSIDER issuance of POTASSIUM IODIDE in accordance with the State Plan Completed by (SED)______________ Peer Checked by ________________

WBN General Emergency EPIP-5 Unit 0 Rev. 0043 Page 30 of 34 Appendix H (Page 1 of 2)

Initial - Protective Action Recommendations

WBN General Emergency EPIP-5 Unit 0 Rev. 0043 Page 31 of 34 Appendix H (Page 2 of 2)

Initial - Protective Action Recommendations

WBN Site Area Emergency EPIP-4 Unit 0 Rev. 0038 Page 15 of 28 Appendix A (Page 1 of 1)

Site Area Emergency Initial Notification Form

1. This is a Drill This is an Actual Event - Repeat - This is an Actual Event
2. The SED at Watts Bar has declared a Site Area Emergency
3. EAL Designator: ________________________________
4. Radiological Conditions: (Check one under both Airborne and Liquid column.)

Airborne Releases Offsite Liquid Releases Offsite Minor releases within federally approved Minor releases within federally approved limits1 limits1 Releases above federally approved limits1 Releases above federally approved limits1 Release information not known Release information not known (1Tech Specs/ODCM) (1Tech Specs/ODCM)

5. Event Declared: Time:________________ Date:__________________

Eastern Time

6. Provide Protective Action Recommendation: None Completed By (SED) _______________________________________

Peer Checked By: _______________________________________

WBN Emergency Plan Classification Logic EPIP-1 Unit 0 Rev. 0042 Page 8 of 54 1.0 PURPOSE This Procedure provides guidance in determining the classification and declaration of an emergency based on plant conditions.

2.0 RESPONSIBILITY The responsibility of declaring an Emergency based on the guidance within this procedure belongs to the Shift Manager/Site Emergency Director (SM/SED) or designated Unit Supervisor (US) when acting as the SM or the TSC Site Emergency Director (SED). The following duties CAN NOT be delegated:

Emergency Classification, Emergency Dose Approval and PAR development prior to CECC Director ownership for PAR development.

3.0 INSTRUCTIONS 3.1 Precautions and Limitations CAUTION Unit-2 radiation monitor readings for classification purposes do not apply until Unit-2 is licensed and operating.

A. The criteria in WBN EPIP-1 are given for GUIDANCE ONLY: knowledge of actual plant conditions or the extent of the emergency may require that additional steps be taken. In all cases, this logic procedure should be combined with the sound judgment of the SM/SED and/or the TSC SED to arrive at a classification for a particular set of circumstances.

B. The Nuclear Power (NP) Radiological Emergency Plan (REP) will be activated when any one of the conditions listed in this logic is detected.

C. The SM/SED shall assess, classify, and declare an emergency condition within 15 minutes after information is first available to plant operators to recognize that an EAL has been exceeded and to make the declaration promptly upon identification of the appropriate Emergency Classification Level (ECL).

1. For EAL thresholds that specify duration of the off-normal condition, the emergency declaration process runs concurrently with the specified threshold duration.
a. Consider as an example, the EAL fire which is not extinguished within 15 minutes of detection. On receipt of a fire alarm, the plant fire brigade is dispatched to the scene to begin fire suppression efforts.

CENTRAL EMERGENCY CONTROL CENTER (CECC) Page 6 of 44 OPERATIONS CECC EPIP-1 Revision 60 3.3 CECC Director/Assistant CECC Director The CECC Director is responsible for directing TVA's overall response to the emergency.

An Assistant CECC Director (who is qualified as a CECC Director) may be used to assist the CECC Director in the accomplishment of position duties. The CECC Director, at his discretion, may delegate the accomplishment of duties to the Assistant CECC Director including signature authority.

The CECC Director ensures that Federal, State, and local agencies are notified in accordance with established procedures and that they are kept fully informed of all aspects of the emergency. The Director reviews with the Plant Assessment and Radiological Assessment Managers the onsite and offsite consequences of the accident and assesses the adequacy and need for measures taken for protection of the public. The Director coordinates TVA's efforts with State and Federal agencies involved in the offsite aspects of the emergency and requests any required federal assistance through the NRC.

Checklists for the CECC Director are provided in Appendices B through G. The CECC Director shall complete Appendix B for initial activation and the appropriate Appendix for the event (e.g.,

Appendix E for an Alert). After the appropriate level of CECC activation the CECC Director is responsible for the following:

  • Approves all press releases developed in the CECC.
  • Notifies the appropriate state warning point of any emergency classification upgrades.
  • Notifies the appropriate state GAR of any emergency classification upgrades.
  • Approves and communicates any required Protective Action Recommendations (PARs) and PAR upgrades to the appropriate state warning point and GAR using Appendix I.
  • Maintains control of Safeguards Information within the CECC.

3.4 Plant Assessment Manager Plant Assessment Manager responsibilities are contained in CECC EPIP-6.

3.5 Radiological Assessment Manager Radiological Assessment Manager responsibilities are contained in CECC EPIP-7.

3.6 Public Information Manager Public Information Manager responsibilities are contained in CECC EPIP-14.

WBN Emergency Exposure Guidelines EPIP-15 Unit 0 Rev. 0016 Page 14 of 16 Appendix C (Page 1 of 2)

Emergency Respirator Issue Guidelines NOTE THESE GUIDELINES ARE RECOMMENDATIONS ONLY, SUBJECT TO THE JUDGEMENT OF RP AND EMERGENCY MANAGEMENT PERSONNEL. THESE GUIDELINES ARE APPLICABLE ONLY TO PROTECTION FROM AIRBORNE RADIOACTIVE MATERIAL AND DO NOT APPLY TO RESPIRATORS/SCBAS ISSUED FOR PROTECTION FROM INDUSTRIAL OR CHEMICAL HAZARDS OR ATMOSPHERES IMMEDIATELY HAZARDOUS TO LIFE OR HEALTH.

TASKS TO SAVE A LIFE OR Respirator/SCBA not required to enter airborne radioactivity PREVENT SIGNIFICANT areas provided resulting internal dose plus external dose will DAMAGE TO PLANT not result in TEDE exceeding NRC dose limits or, if approved by the SED, doses up to the TVA emergency dose limits (i.e., up to 25 Rem/10 Rem) (this can include uptakes > 1 ALI)

HIGH PRIORITY TASKS

  • Respirator/SCBA not required to enter airborne areas if the (priority 1 or 2) following are true:

NOTE: IF INDIVIDUAL'S TOTAL

  • Individual's internal dose plus external dose will not result INTAKE FOR THE YEAR in TEDE exceeding NRC annual dose limit; and TO DATE EXCEEDS 200 DAC-HRS., DOSE
  • Delays or hindrances caused by issuing or wearing RESULTING FROM ALL respirators/SCBAs will jeopardize the success or timeliness INTAKES FOR THE YEAR of the task; or TO DATE MUST BE ASSESSED IN
  • Use of a respirator/SCBA will result in a higher TEDE to the DETERMINING THE TEDE. responding individuals.

LOW or MID PRIORITY TASKS Use RCI-120 for respirator issue guidance.

NOTE Protective requirements may be revised at the discretion of the TSC RP Manager as sample data becomes available.

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

C. Facility licensee procedures required to obtain authority for design and operating changes in the facility. [10 CFR 55.43(b)(3)]

Some examples of SRO exam items for this topic include:

  • Administrative processes for the installation of temporary instrumentation.
  • Processes for changing the plant or plant procedures.

Section IV provides an example of a satisfactory SRO-only question related to this topic.

D. Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

[10 CFR 55.43(b)(4)]

Some examples of SRO exam items for this topic include:

  • Process for gaseous/liquid release approvals, i.e., release permits.
  • Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.

Analysis and interpretation of coolant activity, including comparison to emergency plan criteria and/or regulatory limits.

SRO-only knowledge should not be claimed for questions that can be answered solely based on RO knowledge of radiological safety principles; e.g., RWP requirements, stay-time, DAC-hours, etc.

E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed. One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.

The applicants knowledge can be evaluated at the level of 10 CFR 55.43(b)(5) by ensuring that the additional knowledge of the procedures content is required to correctly answer the written test item, for example:

Page 6 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

  • Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps.
  • Knowledge of diagnostic steps and decision points in the emergency operating procedures (EOP) that involve transitions to event specific sub-procedures or emergency contingency procedures.
  • Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.

SRO-only knowledge should not be claimed for questions that can be answered solely using systems knowledge; e.g.:

  • how the system works.
  • system flow path.
  • component locations, etc.

SRO-only knowledge should not be claimed for questions that can be answered solely using fundamental knowledge of:

  • the basic purpose, the overall sequence of events that will occur, or the overall mitigative strategy of a procedure.
  • any AOP entry condition.
  • plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.
  • immediate operator actions of a procedure.

Section IV and V of this document provide several satisfactory and unsatisfactory examples of test items related to this 10 CFR 55.43(b)(5) topic.

Page 7 of 16

Clarification Guidance for SRO-only Questions Rev 1 (03/11/2010)

Figure 2: Screening for SRO-only linked to 10 CFR 55.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, Yes RO question flowpath, logic, component location?

No Can the question be answered solely by knowing immediate operator actions? Yes RO question No Can the question be answered solely by knowing entry conditions for AOPs or plant parameters Yes RO question that require direct entry to major EOPs?

No Can the question be answered solely by knowing the purpose, overall sequence of events, or Yes RO question overall mitigative strategy of a procedure?

No The SRO with the command function Does the question require one or more of the following? (US or Shift Manager) will

  • Assessing plant conditions (normal, abnormal, or implement the emergency) and then selecting a procedure or section of a REP. Note that the procedure to mitigate, recover, or with which to proceed RO cannot perform
  • Knowledge of when to implement attachments and this function.

appendices, including how to coordinate these items with procedure steps Yes SRO-only

  • Knowledge of diagnostic steps and decision points in the question EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures
  • Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only Page 8 of 16
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