ML15259A292

Watts Bar 2015-301 Draft SRO Written Exam

ML15259A292
Person / Time
Site:	Watts Bar
Issue date:	09/16/2015
From:	Division of Reactor Safety II
To:	Tennessee Valley Authority
References
Download: ML15259A292 (321)

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

Subsequently, the followi ng 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 Malfunct ions During Pump Operation TI-12.04, User's Guide for Abnorma l and Emergency Operat ing 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 , 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 gi ven for an initial controlled sh utdown, then he would select an appropriate procedure (again because the uni t 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 appl icant did not understand that a trip setpoint were exceeded and that the seal leako ff criteria were exceeded.

B. Correct: As mentioned, this dist ractor lists the correct action fo r the control room staff to take; this is, tripping the reactor when an i mmediate 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 se ction 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. Wh en 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 i mmediate 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 in terpret 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 th e 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 out come. Firstly, the Unit Supervisor could immediately identify that RCP immediate tr ip criteria are met through me morization of the criteria of Attachment 2, "RCP Immediate Shut down 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 Leako ff Flow High." Section 3.3 contains a decision step which selects whet her a controlled shutdown (g iven 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 se al leakoff is not in excess of this value, then the Unit Supervisor would assi gn 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 opport unity for plant shutdown (because seal leakoff were less than 6.0 gpm) t hen 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, 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

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

(#1 seal leakoff flow LOW, AND Standpipe level alarm LIT), 3.5 #3 seal leakoff flow HIGH

(#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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of run time bef ore 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 a llow 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. 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:

• AOI-39, Rapid Load Reduction.

• GO-4, Normal Power Operation.

• GO-5, Unit Shutdown From 30%

Reactor Power to Hot Standby.

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 bas ed 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 shutdown required:

• 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:

• Recommendations for continued RCP operation.

• 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 establis hed 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:

• 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:

• VENT VCT by controlling 1-FCV-62-125, OR

• 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. 1 seal outlet temp:

• 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 fo llowing setpoints will require an immediate pump shutdown. Operating limits can be found in SOI 68.02. This list is immediate shut down 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 great er 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 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 ac tions. 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 applie d --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-in terfere basis) with the EOIs. 2. When running an AOI concurrently wit h 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 stat us as required by the AOI. 3. When an AOI in effect directs a Reac tor Trip, then the performance of the AOI should continue immediately fo llowing transition to 1-ES-0.1. Performance assignments will be at t he discretion of the SM/US based on the status and importance of events in progress. 4. When implementing an AOI outside t he "horseshoe" in the control room, the Unit Supervisor should accom pany 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. 1 seal leakoff flow LOW, AND Standpipe level alarm DARK, 3.4 #2 Seal Leakoff Flow HIGH

(#1 seal leakoff flow LOW, AND Standpipe level alarm LIT), 3.5 #3 seal leakoff flow HIGH

(#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 limit s 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.

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 brie f on Steps 3 through 6 prior to tripping the reactor. This ensures that the affected RCP is stopped and that appropriate actions are ta ken 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:

• 1-FR-62-24 [RCP 1 & 2]

• 1-FR-62-50 [RCP 3 & 4]

• 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-62-22 [RCP 2]

• 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.

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 brie f on Steps 10 through 13 prior to reducing load. This ensures t hat the affected RCP is stopped and that appropriate actions are taken w hen 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:

• AOI-39, Rapid Load Reduction

• 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:

• 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 contro l 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:

• 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:

• 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 loop's pressurizer spray valve.

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:

• LCO 3.4.1, RCS Press, Temp and Flow DNB Limits.

• LCO 3.4.2, RCS Minimum Temp For Criticality.

• LCO 3.4.4, RCS Loops Modes 1 and 2.

• LCO 3.4.5, RCS Loops Mode 3.

• 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:

References:

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.

• Administrative processes for temporary modifications.

• Administrative processes for disabling annunciators.

• 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 procedure's overall mitigative strategy or purpose.

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) Page 7 of 16

• 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.

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)

Can the question be answered solely by knowing "systems knowledge", i.e., how the system works, flowpath, logic, component location? RO question Yes NoCan 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 re quire direct entr y to ma jor EOPs?Yes NoDoes 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

• 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 miti gative strate gy of a procedure?No No YesQuestion might not be linked to 10 CFR 55.43(b)(5) for SRO-only NoRO question RO question RO question

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 comp letes 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.

Technical

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:

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 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:

• 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.

• 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.

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.

STEP DESCRIPTION TABLE FOR FR-S.l STEP: Verify Turbine Trip PURPOSE: To ensure that the turbine is tripped BASIS: Step__2__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 HFRSIBG.doc 77 HP-Rev.2, 4/30/2005 STEP DESCRIPTION TABLE FOR FR-S.l STEP: Check AFW Pumps Running PURPOSE: To ensure AFW pumps are running BASIS: Step 3__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 HFRSIBG.doc 79 HP-Rev.2, 4/30/2005 STEP DESCRIPTION TABLE FOR FR-S.l STEP: Initiate Emergency Boration of RCS Step 4__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/orsafeguards 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 HFRSIBG.doc 80 HP-Rev.2, 4/30/2005 ACTIONS: STEP DESCRIPTION TABLE FOR FR-S.l Step 4__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 HFRS1BG.doc 81 HP-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:

• 10 CFR 50.59 screening and evaluation processes.

• Administrative processes for temporary modifications.

• Administrative processes for disabling annunciators.

• 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 procedure's overall mitigative strategy or purpose.

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) Page 7 of 16

• 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.

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)

Can the question be answered solely by knowing "systems knowledge", i.e., how the system works, flowpath, logic, component location? RO question Yes NoCan 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 re quire direct entr y to ma jor EOPs?Yes NoDoes 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

• 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 miti gative strate gy of a procedure?No No YesQuestion might not be linked to 10 CFR 55.43(b)(5) for SRO-only NoRO question RO question RO question

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 chargi ng 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 i ndicated on a status tree, then a transition to that tree's restorat ion procedure is no rmally 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 afte r 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 fo r 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

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.

WBN Unit 1 Steam Generator Tube Rupture 1-E-3 Rev. 0003 Step Action/Expected Response Response Not Obtained Page 11 of 47 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 ra te 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.

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 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:

• 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.

• 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:

• Total available feed flow greater than 410 gpm, OR

• At least one S/G NR level greater than 29%

[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.

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:

• CONTROL PZR heaters as necessary.

• 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:

• Vessel head void formation.

• 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.

• • 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:

• 10 CFR 50.59 screening and evaluation processes.

• Administrative processes for temporary modifications.

• Administrative processes for disabling annunciators.

• 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 procedure's overall mitigative strategy or purpose.

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) Page 7 of 16

• 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.

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)

Can the question be answered solely by knowing "systems knowledge", i.e., how the system works, flowpath, logic, component location? RO question Yes NoCan 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 re quire direct entr y to ma jor EOPs?Yes NoDoes 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

• 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 miti gative strate gy of a procedure?No No YesQuestion might not be linked to 10 CFR 55.43(b)(5) for SRO-only NoRO question RO question RO question

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 00:00:01 1-XS-57-96, 125 VITAL BATT BD VOLTMETER SELECTOR is in

position I 00:01:01 1-EI-57-96, VIT BATT BDS VOLTS reads 131.5 VDC 00:11:00 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. 00:11:00 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 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.

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.

• 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).

• 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.

• 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 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) 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.

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)

Can question be answered solely by knowing hour TS/TRM Action? RO question Yes NoCan 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?

• 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)

• Knowledge of TS bases that is required to analyze TS required actions and terminology SRO-only question Yes NoQuestion might not be linked to 10 CFR 55.43(b)(2) for SRO-only No No No)Knowledge of TS bases that is required to analyze TS grequired actions and terminolog y YesSRO-only quest i on DC Sources - Operating

3.8.4 Watts

Bar-Unit 1 3.8-24 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.

---

NOTES------------------------------------------------

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 DG s 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 power subsystem inoperable.

A.1 Restore vital DC electrical power subsystem to OPERABLE status.

2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 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 power subsystem inoperable.

C.1 Restore DG DC electrical power subsystem to OPERABLE status.

2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (continued)

Nomentionofthechargerismadeinthe"abovethe line"portionoftheT/S.

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.

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)

TheS/Rsarethesourceofthevoltage requirements.

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 du ring 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 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).

An OPERABLE vital DC electrical power subsystem requires all required batteries and respective chargers to be operating and connec ted 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.

Four 125V vital DC electrical power subs ystems, each vital subs ystem channelconsistin g of a batter y bank, asso ciated batter y char ger and the correspondin g contro l equ ipment an d i nterconnect i n g ca bli n g supplyi n g power to t he assoc i ate d D C bus within the c h annel;An OPERABLE vital DC electrical power subsystem requires all required batteries and respective char g ers to be operatin g and connected to the associated DC buses.Notethatachargermustbeoperating andconnected.

80. Given the following conditions: - A perturbation occurred on the 161kV transmission grid. - During the perturbation, the control power supply breaker, fo r 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 descr ibes the operability of the offs ite power supply AND how the CSST C - Y winding voltage will be maintained?

In accordance with T/S LCO 3.8.1, AC Source s - 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 Dispat cher (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 Dispat cher (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 pl aced on the SDBD to maintain it operable. It is plausible to believe this as if such were done, the voltage of the SDBD woul d be certainly maintained within limits. Also, it very reasonably seems counterintuitiv e 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 pl aced 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 Specif ication 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 adequat e 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 operabili ty 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 r equirement was met and that the operability of the offsite source was not maintained.

D. Incorrect: While it is corre ct that a notificatio n 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 ma y 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 shutdow n board. Notice that the allowable voltage range of 6800-7260 VAC is 100 VAC less than 6900VAC and 360 VAC greater than 6900VAC. Theref ore, 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.

Technical

Reference:

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

be provided:

None 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 Unit 1 Main Control Room Voltage Monitoring 1-PI-OPS-1-500KV Rev. 0007 Page 7 of 13

5.1 Voltage

Control Mo nitoring (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 limit ations imposed by the WBN grid operating guide.

Technical Specification operability is maintained in alternate alignment configuration for CSST Load Tap Changers by ensuring trans mission alignments (TRO-TO-SOP-30.130, Watts Bar Nuclear Plant Grid Operating Gu ide) 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 Boar d 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 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 circui t has the correct breaker alignment and indicated power available. B. Each DG tested is capable of starti ng from standby condition or modified start and achieving steady state voltage of great er than or equal to 6800 Volts and less than or equal to 7260 Volts and frequen cy 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) (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.

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.

• 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).

• 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.

• 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 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. 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)

Can question be answered solely by knowing 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action? RO question Yes NoCan 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?

• 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)

• Knowledge of TS bases that is required to analyze TS required actions and terminology SRO-only question Yes NoQuestion might not be linked to 10 CFR 55.43(b)(2) for SRO-only

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 dire ct 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 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."

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 impl emented, 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 beginni ng 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 identi fied in a CAUTION or NOTE at the beginning of the specific instruction. 2. 1-ECA-1.1, Loss of RHR Sump Reci rculation, direct s the operator to perform actions which are intende d 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 cont ainment spray pumps in service. c. The guidance of 1-ECA-1.1 take s 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 conf lict and its correct resolution.

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 Depressu rization 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 prio rity 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 lo ss of shutdown power during which the actions of a FR in effect could NOT be completed successfully. a. If a complete loss of shutdown po wer 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. St atus Tree conditions should be reevaluated after that poi nt in the instruction. 5. 1-ES-1.3, Transfer to Containmen t 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, transiti on to 1-ES-1.3 is appropriate. b. 1-ES-1.3 transfer sequence steps ar e identified by a number on the control board (e.g. #1) to ensure mi nimum flowpath prior to continuing with the instruction in effect. 1-ES-1.3 should be implemented and completed through the transfer seque nce (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 oper ating 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 Tec hnical Support Center.

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 Reci rculation, 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

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-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 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 monito red 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 t he highest priority ORANGE path Function Restoration Instruction will be implemented. F. Once implemented because of any RE D or ORANGE path, that Function Restoration Instruction will be performed to co mpletion 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.

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.

• Administrative processes for temporary modifications.

• Administrative processes for disabling annunciators.

• 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 procedure's overall mitigative strategy or purpose.

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) Page 7 of 16

• 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.

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)

Can the question be answered solely by knowing "systems knowledge", i.e., how the system works, flowpath, logic, component location? RO question Yes NoCan 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 re quire direct entr y to ma jor EOPs?Yes NoDoes 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

• 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 miti gative strate gy of a procedure?No No YesQuestion might not be linked to 10 CFR 55.43(b)(5) for SRO-only NoRO question RO question RO question

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 isstu ck at 215 steps withdrawn. Which ONE of the following descr ibes the appropriate response?

In accordance with ES-0.1, an imm ediate boration __

__(1)____ required.

Assuming that a condition exists where an Immediate Boration IS required, the SRO will assign the responsibility for the perform ance 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 st ep would lead one to believe that action would be required on the account of one r od 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 acti on (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 fa ilure of a rod to insert).

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 oper ator responsibility for the AOI."

Therefore, it is not correct that the OA C 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. borat e 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). Ther efore, TI-12.04 re quires that the OAC and BOP/CRO perform a "role swap" in this specific instance.

B. Incorrect: As described, it is correct t hat the BOP/CRO would be assigned 1-AOI-34. However it is incorrect and yet plaus ible 1-AOI-34 would be implemented on account of one rod whic h 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

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.

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:

• 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:

• 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:

• 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 ac tions. 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 applie d --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-in terfere basis) with the EOIs. 2. When running an AOI concurrently wit h 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 stat us as required by the AOI. 3. When an AOI in effect directs a Reac tor Trip, then the performance of the AOI should continue immediately fo llowing transition to 1-ES-0.1. Performance assignments will be at t he discretion of the SM/US based on the status and importance of events in progress. 4. When implementing an AOI outside t he "horseshoe" in the control room, the Unit Supervisor should accom pany 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:

• 10 CFR 50.59 screening and evaluation processes.

• Administrative processes for temporary modifications.

• Administrative processes for disabling annunciators.

• 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 procedure's overall mitigative strategy or purpose.

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) Page 7 of 16

• 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.

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)

Can the question be answered solely by knowing "systems knowledge", i.e., how the system works, flowpath, logic, component location? RO question Yes NoCan 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 re quire direct entr y to ma jor EOPs?Yes NoDoes 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

• 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 miti gative strate gy of a procedure?No No YesQuestion might not be linked to 10 CFR 55.43(b)(5) for SRO-only NoRO question RO question RO question

83. Given the following conditions: AOI-34, Immediate Boration, Se ction 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 co mpletes the statement below?

In order to borate the RCS, the US must __________. A. place the "C" BAT in service to th e U1 CVCS blender in accordance with 1-SOI-62.05 B. continue in section 3.

2 of 1-AOI-34 and align t he 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 suct ion to and from the RWST Using 1-LCV-62-135 and 136 in acco rdance 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

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."

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 C B 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 isolati on, 1-ISV-62-933

[A4V/713].

WBN Unit 1 Immediate Boration 1-AOI-34 Rev. 0001

3.2 Boration

of RCS with CV CS 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:

• Neutron flux dropping.

• Tavg dropping.

• Control rod bank position rising (if in AUTO).

IF normal boration NOT inserting negative reactivity, THEN

• • 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.

WBN Unit 1 Immediate Boration 1-AOI-34 Rev. 0001

3.2 Boration

of RCS with CV CS 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. 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:

• 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.

• 3.1.7, 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.

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]. ________

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 IV BA 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 IV BA 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 Su ction from the VCT to the RWST is desired, THEN PERFORM the following:

[1.1] OPEN RWST to CVCS Char ging 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 A nd 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:

• 10 CFR 50.59 screening and evaluation processes.

• Administrative processes for temporary modifications.

• Administrative processes for disabling annunciators.

• 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 procedure's overall mitigative strategy or purpose.

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) Page 7 of 16

• 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.

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)

Can the question be answered solely by knowing "systems knowledge", i.e., how the system works, flowpath, logic, component location? RO question Yes NoCan 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 re quire direct entr y to ma jor EOPs?Yes NoDoes 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

• 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 miti gative strate gy of a procedure?No No YesQuestion might not be linked to 10 CFR 55.43(b)(5) for SRO-only NoRO question RO question RO question

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, Nucl ear 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 co mpletes 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 Un it 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 appl icable in Modes 1,2,3 or 4.

Therefore, it is incorrect that wit h 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 per formance 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 reacto r trip in Mode 5. It is also incorrect that the firs t 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 interv al 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 ye t 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, 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.

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 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 t he EOIs assumes that the safety-related equipment required by Tech Specs in M ode 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 spec ific instruction's applicability. 1. The EOI network assumes that the Residual Heat Re moval (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 a nd 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 mitigat ed 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 re sult 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 (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 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 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 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. 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.

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.

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.

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.

• 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).

• 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.

• 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 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. 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)

Can question be answered solely by knowing 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action? RO question Yes NoCan 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?

• 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)

• Knowledge of TS bases that is required to analyze TS required actions and terminology SRO-only question Yes NoQuestion 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:

• 10 CFR 50.59 screening and evaluation processes.

• Administrative processes for temporary modifications.

• Administrative processes for disabling annunciators.

• 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 procedure's overall mitigative strategy or purpose.

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) Page 7 of 16

• 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.

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)

Can the question be answered solely by knowing "systems knowledge", i.e., how the system works, flowpath, logic, component location? RO question Yes NoCan 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 re quire direct entr y to ma jor EOPs?Yes NoDoes 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

• 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 miti gative strate gy of a procedure?No No YesQuestion might not be linked to 10 CFR 55.43(b)(5) for SRO-only NoRO question RO question RO question

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 bas is 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 ma y 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 Westi nghouse basis document for FR-H.2, Steam Generator Overpressure, "If the level is grea ter than [93%], the SG water level may be above the narrow range or the SG may even be filled solid with water."

Therefore, the val ue observed on ICS (the computer screen) is gr eater 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 potentia lly 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 Emer gency Operating Instructions,"

"During performance of the EOI set, the operator is required to utilize PAM instruments when they are pr ovided on the control board."

Therefore, as the 1-M-4 indicati ons 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 gr eater than 93%; therefor e, it is incorrect to believe that all indications relate that the SG is potentially filled solid with water. It would be correct th at 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 IC S data suggests that the SG is potentially filled solid, it is not true (as previ ously 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 pos t-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 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.

Technical

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 requi red 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]

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.

ACTIONS: 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 HFRH2BG.doc 11 HP-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 t heir designator in bold face type. 1. The designator is followed by the text extending across the entire page with note text appearing in standard ty pe and caution text appearing in bold face type. 2. If multiple cautions or notes are applicable to a step t hen each caution or note included after the initial designator is disti nguished 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 co mmunicated (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) instrum entation is provided on the main control board as determined by design requirements. 1. The control room complies with Reg.

Guide 1.97 requirem ents by providing the operator with the requi red PAM instruments. 2. The control board PAM instruments ar e uniquely labeled to identify them as PAM instrumentation. a. Most PAM instruments have black background instrumentation labels. b. Other PAM instruments are identif ied 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 pr ovided on the control board. 4. The operator should compare redundant instruments when they are provided. 5. Some parameters evaluated dur ing performance of the EOIs do NOT have PAM grade instrumentation pr ovided on the control board. a. The operator should monitor th ese parameters with available instrumentation.

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:

• S/G MFW isolation and bypass isolation valves CLOSED.

• S/G MFW reg and bypass reg valves CLOSED.

• MFP A and B TRIPPED.

• Standby MFP STOPPED.

• Cond demin pumps TRIPPED.

• Cond booster pumps TRIPPED.

Manually CLOSE valves, AND STOP pumps, as necessary.

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:

• OPEN S/G PORVs, OR

• OPEN S/G MSIV bypass valves, OR

• OPEN S/G steam supply to TD AFW pump, OR

• 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:

• 10 CFR 50.59 screening and evaluation processes.

• Administrative processes for temporary modifications.

• Administrative processes for disabling annunciators.

• 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 procedure's overall mitigative strategy or purpose.

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) Page 7 of 16

• 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.

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)

Can the question be answered solely by knowing "systems knowledge", i.e., how the system works, flowpath, logic, component location? RO question Yes NoCan 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 re quire direct entr y to ma jor EOPs?Yes NoDoes 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

• 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 miti gative strate gy of a procedure?No No YesQuestion might not be linked to 10 CFR 55.43(b)(5) for SRO-only NoRO question RO question RO question

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.

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 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.

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.

• 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).

• 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.

• 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 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) 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)

Can question be answered solely by knowing hour TS/TRM Action? RO question Yes NoCan 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?

• 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)

• Knowledge of TS bases that is required to analyze TS required actions and terminology SRO-only question Yes NoQuestion might not be linked to 10 CFR 55.43(b)(2) for SRO-only No No No* )Knowledge of TS bases that is required to analyze TS grequired actions and terminolog y YesSRO-only quest i onTherefore,thedeterminationof operabilityofthe trainofSSPSis SROonly.

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.

• Administrative processes for temporary modifications.

• Administrative processes for disabling annunciators.

• 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:

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.

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.

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.

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.

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.

Automatic Trip Lo gic (Functio n 19) 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 Lo g ic to be OPERABLE. Havin g two OPERABLE channels ensures that r a ndomfailure of a sin g le lo gic channel will not prevent reactor trip

.NotethattheSSPSremainscompletely capableof performingthis functionfollowing thefailureofthe singlepower

supply.

Thepowersuppliesareredundant.

PER Vault Summary Report for PER: 3516 PER Number:

3516 Status: ARCHIVE Status By ID:

Status Date:

Site / Org:

WBN Unit: CAP Due Date: Date of Occurence: Long Lead CA Date: Analysis Type:

PER Level:

PER Summary:

Problem Evaluation Report - User Information Originator 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 Details Initiating Department:

Site / Org:

WBN SBU: 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:

NO Problem Evaluation Report - Customer ImpactOutage

Reference:

Customer

Reference:

As Found Condition:

Problem Evaluation Report - Regulatory Impact Potential Environmental Issue:

NO Potential Safety Issue:

NOPotential Operability Issue: Potential Reactivity Issue:

Potential Reportability Issue:

NO NO PER Level:

PER Category: Tier: Management Screening - Review Results / Approval Good Catch: Analysis Type: CAP Due Date:

Responsible Org:

Comments:

Justification Details:

Comment Details:

NO Potential Margin Issue:

NO Justification:

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

,, T R SS P S I S STILL O PERABLE D U E T O THE A UC TI O NEERIN G C IR CUIT F OR THE 1 5 VD C PWR SU PPLIE S.

PER Vault Summary Report for PER: 3516Event Number: Event

Description:

Regulatory Reviews - Environmental ReviewEnvironmental Issue?: Environmental Type:

Environmental Area:

EMS Process:

Source Code:

Findings Code: Notice of Violation?:

NO Event Repeat?:

NO Justification for Environ. Disposition:

Reviewer ID:

Review Date:

Non-Conformance Code:

Justification Details:

Potential Environmental Issue?:

NO Regulatory Reviews - Safety ReviewSafety Event Number: Potential Safety?:

NO Safety Issue?:

Source Code:

Findings Code: Violation Notice Required?:

NO Justification for Safety Disposition:

Reviewer ID:

Review Date:

Justification Details:

Regulatory Reviews - Operations Review Potential Operability Issue?:

Operability Issue?:

Operability Actions: Operability Actions Details:

Potential Reportability Issue?:

Reportability Issue?:

Operability Reviewer:

Operability Review Date:

Reportability Reviewer:

Reportability Review Date:

NO NO NO NO Engineering Evaluation Needed?:

Required Date:

Ops Notified Other Sites?:

NO Regulatory Reviews - Engineering Evaluation Specified (Safety) Function Maintained?:

Evaluation Summary:

CLB Affected: CLB Affected Details:

Immediate/Compensatory Measures?:

NO NO Engineering Evaluation Needed?:

Required Date:

Evaluation Details:

Outside CLB?:

NO Page 3 of 25Wednesday, December 10, 2014TVA RESTRICTED INFORMATIONThecrewscreenedtheissue (correctly)asnot animpacttothe operabilityof

SSPS.Potential Operability Issue?

NO

Notanimpactto operability.

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.

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.

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.

87. Given the following conditions: - Unit 1 is in MODE 3.

Subsequently: - A Loss of 120v AC Vital Instru ment Power Board 1-III occurs. AOI-25.03, Loss of 120V AC Vital Inst rument Power Boards 1-III or 2-III, directs use of 1-SOI-235.03, 120V AC Vi tal Power System 1-III, to restore the board Which ONE of the following descr ibes 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, Invert ers and 3.8.9, Distributi on Systems - Operating to be exited.

A. (1) ONLY the master (2) 8.1, Transferring 480V AC Vital Tr ansfer Switch III to Alternate 480V Power Supply B. (1) BOTH the master and the slave (2) 8.1, Transferring 480V AC Vital Trans fer Switch III to Alternate 480V Power Supply C. (1) ONLY the master (2) 8.3, Transfer 120V AC Vital Inst rument 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 Inst rument 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 ma y 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 th is bus's power is the operating coils for the SSPS master re lays. 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 pow er 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 t he 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 pow er from an inverter unit. This unit can provide power fr om 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 in verter 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 afor ementioned sources of power to supply a 120V AC bus. One may fu rther 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 plausib ility 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 a nd 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 plausib le that only the mast er 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, bot h 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 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)

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."

Bus I Bus II Bus III Bus IVTo SlaveRelaysTo SlaveRelays SSPS Power Supply DistributionTrain ATrain B120V AC Vital Instrumentation 48v 15v 48v 15v I II III IV I II III IV 48v 15v 15v 48vFrom 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."

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.

(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 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 ha ve 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."

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 SHUT DOWN 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 engineeri ng evaluation for this performance and attach a copy to this Data Package. ________ SRO [2] CHECK 1-EI-235-3/V2, BATTERY INPU T on Inverter 1-III OR 0-EI-235-3/V2, BATTERY INPUT on In verter 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 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 f eeds from 120V Instrument Power Distribution Panel 2-A, including feeds from Panel 2-M-7, due to momentary lo ss 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 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. ________

[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 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 INPU T on Inverter 1-III OR 0-EI-235-3/V2, BATTERY INPUT on In verter 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 f eeds from 120V Instrument Power Distribution Panel 2-A, including feeds from Panel 2-M-7, due to temporary lo ss 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 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. ________ ________

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 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 transfe rs 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 SU PPLY 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.

• 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).

• 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.

• 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 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. 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)

Can question be answered solely by knowing 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action? RO question Yes NoCan 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?

• 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)

• Knowledge of TS bases that is required to analyze TS required actions and terminology SRO-only question Yes NoQuestion might not be linked to 10 CFR 55.43(b)(2) for SRO-only

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.

3. 1-LCV-3-164, MDAFWP SG 1 SUPPLY.

Of the THREE equipment items list ed above, which ONE of the foll owing 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 aut omatically start up on 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 emer gency exhaust fan is safety-related.

The basis for T/S LCO 3.7.5 reflects t hat: "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, t he 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 INOPERAB ILITY renders T/S LCO 3.7.5 NOT MET, it is not correct that the AC exh aust 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 a ccordance 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 emer gency 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 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:

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.

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.

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)

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 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

°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) 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.

• 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).

• 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.

• 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 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. 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)

Can question be answered solely by knowing 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action? RO question Yes NoCan 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?

• 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)

• Knowledge of TS bases that is required to analyze TS required actions and terminology SRO-only question Yes NoQuestion might not be linked to 10 CFR 55.43(b)(2) for SRO-only

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, sect ion 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 TRACKI NG ONLY LCO for the ERCW system C. a train of ERCW is INOPERAB LE 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 INOPERAB LE 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 shutdow n 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 dies el generators" indicates the T/S impact of the emergency power sele ctor 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 theref ore, that ERCW pump which is selected by the emergency power selector switch.

In the conditions depicted in the st em of the question, the A-A ERCW pump has tripped. The buzzer, wh ite and green indicating lights and lack of pressure indication all re late that the pump had been running and is now tripped. The emergency power selector switch can be seen in the A-A pump position. T herefore, 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 t he ERCW train will only be restored OPERABLE when the A-A ERCW pump is repaired and rete sted. 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 woul d 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, t hen the applicant would arrive at this distractor. Specifically, take the example of the Unit 1, "A" train CCS. If the 1B CCS pump were r unning (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 acti on required to restor e the operability of the "A" train ERCW is to reposit ion 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. Technical

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

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 (continued)

Watts Bar-Unit 1 B 3.7-43 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, 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 othe r 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 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 (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 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

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.

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.

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 VALVE A-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:

• 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.

• 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.

• 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 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. 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)

Can question be answered solely by knowing 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action? RO question Yes NoCan 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?

• 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)

• Knowledge of TS bases that is required to analyze TS required actions and terminology SRO-only question Yes NoQuestion might not be linked to 10 CFR 55.43(b)(2) for SRO-only

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 posit ioned 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 44 and the impact on Containment Spray actuation?

1-PDT-30-43 is required to be placed in the ____(1)____ posit ion 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 be provided:

None 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 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 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 54 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br /> 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> (continued)

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 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. OR E.2.1 Be in MODE 3. AND E.2.2 Be in MODE 4. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> 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 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 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 54 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br /> 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> (continued)

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 3.3.2.1 0 (continued)

(a) Above the P-11 (Pressurizer Pressure) Interlock.

(b) Time constants used in the lead/lag controller are t 1 50 seconds and t 2 5 seconds.

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 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 MODE S 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) ps i 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 inch es 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 t 1 50 seconds and t 2 5 seconds. (c) Except when all MSIVs are closed a nd 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 t 3 and t 4 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.

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 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 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 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 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.

• 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).

• 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.

• 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 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. 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)

Can question be answered solely by knowing 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action? RO question Yes NoCan 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?

• 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)

• Knowledge of TS bases that is required to analyze TS required actions and terminology SRO-only question Yes NoQuestion might not be linked to 10 CFR 55.43(b)(2) for SRO-only

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°F slowly RISING - ALL Power Range Detectors indicate 4%. - ALL Intermediate Range SUR Monitors indicate -0.1 dpm.

Which ONE of the following co mpletes the statements below?

At 00: 1 5:00 , T (T = twall - t cc), as shown in the picture abov e, 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 Cool ing Studies of Scenarios With Feedwater Available, Using the NOT RUMP 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 meta l temperature it was related to such.

As seen in the Westinghouse background docum ent 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 elec trically heated [rod] during a critical heat flux transient," demonstrates that once the heat flux present within a process flowpath increases such that nucleate boili ng 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 te mperature of approximately 660°F for steady state operation at rated po wer 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. T herefore, the nominal difference between the hottest fuel rod's metal temperat ure and that register ed 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 st artup 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 transit ion 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 ac tions 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 tran sition 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

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.

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°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°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 DECISION: BLOCK DESCRIPTION TABLE FOR STATUS TREE F-0.2 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:

Core exit thermocouples temperature indication KNOWLEDGE:

N/A 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 HF02BG.doc 6 HP/LP-Rev.

2, 4/30/2005 Thermalhydraulic Behavior of Electrically Heated Rod during a Critical Heat Flux Transient Rita de Cássia Fernandes de Lima Mechanical Engineering Department. Un iversidade 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 radioa ctive 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 theoretica l analysis of the drying and rewetting front propagation during a critical heat flux experiment, starting with the application of an el ectrical 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 analytic ally observed. Studies performed with various values of coolant mass flow rate show that this variable has more influence on the drying front veloci ty 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 oc currence 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 ph enomenon, it is a common practice to perform simulations of the reactor oper ational and transien t conditions in thermalhydraulic loops, utilizing electr ically 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-establishe d 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 & Ya ckle (1981) establish the difference between quench and rewetting. Yu et alii (1977) analyze the quench process on hot surfaces using bidimensiona l conduction. The work treats the subcooled and saturated rewetting for pr essures from 1 to 69 bar. Olek et alii (1988) study the rewetting in de scendent 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 se ction, 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 c p , the specific heat of each materi al. 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 h f , 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; T f is the coolant temperature and R ext =R 4 , 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 fo rmulation. 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 , h ref 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 q o~ (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. Th e 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 intern al 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 o ccurs 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 th ree 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 increa se 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 beha vior can be explaine d by the reduction of the superficial flux and consequently a smaller increase of the enthalpy.

The maximum flow quality in this transi ent is equal to 0.57 at the outlet of the channel at t = 4.0 s.

The effect of the heat transfer regime s 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 1100 o C in less than 7.0 s.

The propagation of the drying and re wetting 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 o C. A 41% reduction in the pressure value has a strong influence on the rewe tting front velocity: an increase of 936%. Otherwise there is a little influenc e 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 ra te has a smaller effect in the drying front. For a 20% reduction in mass fl ow 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 prop agation velocity for the drying out and rewetting processes, during the o ccurrence of critical heat flux in electrically heated simulators of nucle ar fuel rods, caused by a power step.

This study is very importan t 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 critic al 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 de lay 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 rewe tting 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 ra tes 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 11 8° C/s, which indicates that the wall temperature would reach its temperat ure limit, estimated around 1100° C, in approximately 7s. This is the amount of time available to turn off the electrical power supply. This observed he ating rate is much larger than the value obtain by Mosaad (1988).

Additional details of the present wo rk 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 conducti vity; De , the hydraulic diameter of the channel; Re, the Reynolds number; and Pr , the Prandtl one.

b) Nucleate boiling: THOM~s corr elation (TONG & WEISMAN, (1979)).

where Twall = external temperature of the cladding; T sat = saturation temperature; q" SUP = heat flux; e p = pressure.

The transition between forced conv ection 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 conv ection 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 h ref 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 T HN is the homogeneous nucleation temperature. This is the temperature at which the nucleation o ccurs spontaneously in the liquid in the absence of preferred nucleation sites. It is function of the pressure and can be predicted using standard nucleatio n theory. It can be obtained by the following expression of the TRAC -PF1 handbook:

where P = 3203.6 - P, ( in psia) and T HN in 0 F. f) Film boiling: modified Groeneve ld~s correlation (BJORNARD, (1977)).

where: Y = two-phase flow factor of Miropol~skiy, x = flow quality; Pr wall = Prandtl number evaluated at temperature T wall; 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; k g.= thermal conductivity -gaseous phase.

References BJORNARD, T. A.; GRIFFITH, P. PW R 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 De sign, 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 transf er to flowing water in a vertical tube. Doctoral Thesis, Technisch en Universitaet Berlin , 1988.

[ Links ] OLEK, S., ZVIRIN, Y. & ELIAS, E., "Rewe tting 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 4.4-7 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°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, F Q, 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 F Q 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 UO 2 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 Secti on 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°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.

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:

• 10 CFR 50.59 screening and evaluation processes.

• Administrative processes for temporary modifications.

• Administrative processes for disabling annunciators.

• 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 procedure's overall mitigative strategy or purpose.

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) Page 7 of 16

• 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.

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)

Can the question be answered solely by knowing "systems knowledge", i.e., how the system works, flowpath, logic, component location? RO question Yes NoCan 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 re quire direct entr y to ma jor EOPs?Yes NoDoes 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

• 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 miti gative strate gy of a procedure?No No YesQuestion might not be linked to 10 CFR 55.43(b)(5) for SRO-only NoRO question RO question RO question

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 approac hes 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, woul d 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 Ga s Treatment System T/S LCO 3.6.4, Containment Pressure CORRECT ANSWER:

D DISTRACTOR ANALYSIS:

A. Incorrect: As mentioned in the (D) distract or 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 EG TS system is not correct. As previously discussed, EGTS would assist in controlling containment pressure ONLY if the two dam pers 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 volume s would be cross-connected.

Therefore, choosing such distractor is plausible but incorrect as the containment pressure relief valves we re 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 implicati on is present that such procedure would enable the containment purge to be accomplished.

D. Correct: As seen in system descrip tion WBN-SDD-N3-30RB-4002, "Reactor Building Ventilation System:"

"The containment venting, for continuous pressure relief, is perfo rmed during modes 1-5, by opening the containment isolation (CI) valv es 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 contai nment will uncontrollably rise. As seen in system description N3 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 is olation (phase A) is received.

Furthermore, the flowpath through t he EGTS subsystem shows that air is drawn from the annulus, proc essed 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 bu ilding exhaust stack. This division of discharge maintains the annulus at a slight vacuum (relative to the isolated containm ent 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 annotat ion 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 t he 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 contai ned 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 im plemented by the Offsite Dose Instructions (ODIs). Of import to this question are two ODIs. 1-ODI-90-15, "Containment Purge Release" satisf ies 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 th rough the Unit's shield building stack. The outputs of t hese ODIs are the releas e permits. Therefore, it is correct that before containmen t 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 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

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 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

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 eit her 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 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

°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 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 Gases 2 (Gamma Emitters) 1x10-4 H-3 (oxide) 1x10-6 B. Containment PURGE 3 P 8 Each PUR GE Grab Sample P Each Purge Noble Gases 2 (Gamma Emitters) 1x10-4 C. Incore Instrument Room PURGE 3 Each PURGE 9 Grab Sample Each Purge Noble Gases 2 (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 Gases 2 (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 Continuous 6 Tritium Sample W H-3 (oxide) 1x10-6 Continuous 6 Charcoal Sample W 7 I-131 I-133 1x10-12 1x10-10 Continuous 6 Particulate Sample W 7 Principal Gamma Emitters 2 1x10-11 Continuous 6 Composite Particulate Sample M Gross Alpha 1x 10-11 Q Sr-89, Sr-90 1x10-11

• 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 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) 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.

• 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).

• 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.

• 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 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. 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)

Can question be answered solely by knowing 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action? RO question Yes NoCan 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?

• 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)

• Knowledge of TS bases that is required to analyze TS required actions and terminology SRO-only question Yes NoQuestion 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:

• 10 CFR 50.59 screening and evaluation processes.

• Administrative processes for temporary modifications.

• Administrative processes for disabling annunciators.

• 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 procedure's overall mitigative strategy or purpose.

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) Page 7 of 16

• 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.

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)

Can the question be answered solely by knowing "systems knowledge", i.e., how the system works, flowpath, logic, component location? RO question Yes NoCan 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 re quire direct entr y to ma jor EOPs?Yes NoDoes 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

• 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 miti gative strate gy of a procedure?No No YesQuestion might not be linked to 10 CFR 55.43(b)(5) for SRO-only NoRO question RO question RO question

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 descri bes which document for which the MINIMUM Flow is a basis AND the method for providing ma keup 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 sect ion 8.3.1 of 0-SOI-24.01 B. (1) the NPDES permit for WBNP (2) ONLY the RCW system using sect ion 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 secti on 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 secti on 8.3.1 of 0-SOI-24.01 NOTE: SOI-27.03, Supplement al Condenser Circul ating Water System 0-SOI-24.01, Raw Co oling 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.

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 t hat, "The UHS shall be OPERABLE." The basis for this LCO indicates: "T he UHS is required to be OPERABLE and is considered OPERABLE if it contains water at or below the maximum temperature that would a llow 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°F." The basis for this T/S does not menti on 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 dur ing 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." Ther efore, 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 NPDE S permit is the source for the step [1] cited from Section 7.1, "Shu tdown" 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. T herefore, it is incorrect that this mode of cooling tower makeup would be utilized (because the Unit 2 cooling tower is drained for maintenanc e). 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 origin al plant design (e.g. before the installation of SCCW). The operati ng crew could divert some of the RCW supply directly to the cooling to wer flumes to provide makeup to the cooling tower basins and thus main tain 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 app licant must select the correct SOI section in order to provi de 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

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.

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 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 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 85F. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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.

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

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 (coolwater)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).**

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.

• 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).

• 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.

• 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 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. 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)

Can question be answered solely by knowing 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action? RO question Yes NoCan 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?

• 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)

• Knowledge of TS bases that is required to analyze TS required actions and terminology SRO-only question Yes NoQuestion might not be linked to 10 CFR 55.43(b)(2) for SRO-only

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 ac tion 3.4.16 A.1, DEI must be verified

< ____(1)____ µCi/gm.

The NOTE "LCO 3.0.4.c is applicabl e" indicates that when an LCO is not met, entry into a MODE or other specified condition in the Applic ability 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 assessm ent 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 asse ssment addressing in operable 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 no t 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 ri sk assessment addressing inoperable systems and components, considerati on of the results, determination of the acceptability of entering the MODE or ot her specified condition in the Applicability, and establishm ent of risk management actions, if appropriate; exceptions to this Specif ication are stated in the individual Specifications, or

c. When an allowance is stated in t he 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 acti on A.1 for T/S LCO 3.4.16, a verification must be made that DEI is <14 µ 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 µCi/gm. It is plausible to believe t hat 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 dur ing the SRO applicant's time in initial license training.

It is Correct: however, that a risk a ssessment 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 t hat 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 r equired 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 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:

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 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; 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.

RCS Specific Activity 3.4.16 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.

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

> 0.265 Ci/gm.

---

NOTE-------------------- LCO 3.0.4.c is applicable. ---------------------------------------------------

A.1 Verify DOSE EQUIVALENT I-131 14 Ci/gm AND A.2 Restore DOSE EQUIVALENT I-131 to within limit.

Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />

48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 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 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 6 hours (continued)

RCS Specific Activity 3.4.16 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 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

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 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)

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 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 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 />. 184 days RCS Specific Activity 3.4.16 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 t he reactor coolant shall be within limits.

APPLICABILITY: MODES 1 and 2, MODE 3 with RCS average temperature (Tavg) 500 F.

ACTIONS CONDITION REQUIRED ACTI ON 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 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />

48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 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 < 500 F. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (continued)

RCS Specific Activity 3.4.16 Watts Bar-Unit 1 3.4-40 Amendment 41

ACTIONS (continued)

CONDITION REQUIRED ACTI ON COMPLETION TIME C. Required Action and associated Completion Time

of Condition A not met.

OR DOSE EQUIVALENT I-131 >

21 Ci/gm. C.1 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 />

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 Verify reac tor 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 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)

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.

• 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).

• 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.

• 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 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. 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)

Can question be answered solely by knowing 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action? RO question Yes NoCan 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?

• 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)

• Knowledge of TS bases that is required to analyze TS required actions and terminology SRO-only question Yes NoQuestion might not be linked to 10 CFR 55.43(b)(2) for SRO-only

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: "Equip ment removed from service or declared inoperable to comply with ACTION S may be returned to service under administrative control solely to per form testing required to demonstrate its OPERABILITY or the OPERABILITY

of other equipment." T herefore, this answer is plausible because it is partly correct (i.e. that it is corre ct 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 pr ocess which restores a failed component to an operable status. O ne 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 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:

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.

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.

• 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).

• 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.

• 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 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. 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)

Can question be answered solely by knowing 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action? RO question Yes NoCan 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?

• 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)

• Knowledge of TS bases that is required to analyze TS required actions and terminology SRO-only question Yes NoQuestion might not be linked to 10 CFR 55.43(b)(2) for SRO-only

96. In accordance with NPG-SPP-01.2, Administration of Site Technical Procedures, which ONE of the following describes the r equirements for making a MINOR/EDITORIAL change to a technical procedure?

An Independent Qualified Revi ew (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 r equired 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 screeni ng is not required.

C. Incorrect: NPG-SPP-01.2, Administration of Site Technical Procedures, 3.2.11.A describes that an AOR is not requi red but an IQR is 50.59 screening requirements are direct ed 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

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 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.

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 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.

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.

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

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) 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.

• Administrative processes for temporary modifications.

• Administrative processes for disabling annunciators.

• 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 procedure's overall mitigative strategy or purpose.

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:

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 in side 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 lim it in any core thimble (2) Hold Order D. (1) OR approximately ten feet below the bottom of the core lim it 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 descri bed 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 operat ion 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 lo cations 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 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:

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 t he 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 Po lar 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 wit h the RP Manager) must evaluate the necessity of the entry, issue special inst ructions (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 associ ated 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 implement ation of this TI. The br iefing as a minimum is to cover items contained in Step 3.2.1. C and D.

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 Pr ogram 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 buildi ng 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 applic able sections of Appendix A, Containment/Annulus Entry Authorization. C. One Appendix A is required for eac h area entered (Upper Containment, Lower Containment, Annulus). Appendix A's ar e 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 approv e 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 comple tion 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, Sect ion 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 ent ries 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.

• Administrative processes for temporary modifications.

• Administrative processes for disabling annunciators.

• 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 procedure's overall mitigative strategy or purpose.

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. 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 contai nment entry listed above.

The personnel hazard stipulat ed 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 Wa ll while in Mode 1 or 2-the SM (concurrent with the RP Manager)-authorize the entry by co mpletion 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 A ccess Authorization shall require approval s-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 fo r the containment entry.

It is not correct that the crew woul d 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 contai nment 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, th is 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 m entioned chapter of the safety manual and not those of TI-12.07A. Al so amplifying this plausibility is the fact that TI-12.07B, "Containm ent Access Modes 5&6" does regard heat stress as a concern and ensures that persons conducting closeout inspections of the containm ent (for entry in to 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 c oncern 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 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:

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 Pr ogram 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 buildi ng 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 applic able sections of Appendix A, Containment/Annulus Entry Authorization. C. One Appendix A is required for eac h area entered (Upper Containment, Lower Containment, Annulus). Appendix A's ar e 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 approv e 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 comple tion 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, Sect ion 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 ent ries 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).

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 en try. 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 establis h 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 desig nee. 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 concer n, 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 re sponsible for ensuring completion and submittal of paperwork to Operations for closure. E. All personnel must obtain an RWP brie fing 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 contac ted 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 operat ion 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 utiliz ing Appendix B, Personnel Accountability Logsheet(s).

These logs are maintained in the following manner:

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 t he 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 Po lar 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 wit h the RP Manager) must evaluate the necessity of the entry, issue special inst ructions (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 associ ated 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 implement ation of this TI. The br iefing as a minimum is to cover items contained in Step 3.2.1. C and D.

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.

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.

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:

• personnel safety precautions and protective measures (climbing, heat stre ss, ALARA, etc)

• detailed acceptance criter ia in Section 5.0 for housekeeping/cleanliness st andards 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 oper ation (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) 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.

• Administrative processes for temporary modifications.

• Administrative processes for disabling annunciators.

• 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 procedure's overall mitigative strategy or purpose.

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) Page 7 of 16

• 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.

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)

Can the question be answered solely by knowing "systems knowledge", i.e., how the system works, flowpath, logic, component location? RO question Yes NoCan 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 re quire direct entr y to ma jor EOPs?Yes NoDoes 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

• 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 miti gative strate gy of a procedure?No No YesQuestion might not be linked to 10 CFR 55.43(b)(5) for SRO-only NoRO question RO question RO question

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 implement ation 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 relat ed 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 ther e is a fire in the plant. At 00:0 1:00 , conditions exist which indicate t hat the fire in the SDBD room has caused multiple spurious actuat ion (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 singl e 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 wa it 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 porti on 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 e quipment involved." Then the SRO must decide that a loss of plant c ontrol 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 a fter an entry is made into 1-AOI-30.2 from 1-AOI-30.1 after the crit eria aforementioned are noted.

C. Incorrect: The background seen in 1-AOI-30.2 (pg 12) relates: "I. No other accident is assumed to occur concu rrently with a fire." Therefore, while it is true that the Appen dix R fire does not analyze for subsequent casualties, as mentioned prev iously 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 demons trated the "potential to affect plant control (safe shut down 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

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."

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):

• Multiple spurious actuations of systems/components.

• Erratic or questionable indications on numerous MCR meters/recorders.

• 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

• • GO TO AOI-30.2, Fire Safe Shutdown.

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 decl are an Appendix R fire is left to the judgment of t he Unit SRO/SM and must be based on the magnitude of the fire and its potent ial 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 predominat

e. 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 5.1 E and 5.1 F. For each safety functi on, the equipm ent required to accomplish the safety function has been divided into "Keys" which represent groups of functionally-relat ed 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 coul d 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.

For example, the BIT outlet valves ar e 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 availabl e 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 shutdow n 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 Bu ilding, it is essential that, functionally, the same equipment and instrumentati on be available from the Aux Control Room or remote stations or otherwise be justifi ed. Loss of offsite power is assumed concurrently with MCR fires. K. The possibility of shutdow n 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 oper ation of a component coul d defeat the r equired system safety function, manual actions are tak en to address the effects of spurious component operation. Com ponents identified as thos e which could prevent a safe shutdown should they s purious operate are those that:

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.

• Administrative processes for temporary modifications.

• Administrative processes for disabling annunciators.

• 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 procedure's overall mitigative strategy or purpose.

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) Page 7 of 16

• 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.

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)

Can the question be answered solely by knowing "systems knowledge", i.e., how the system works, flowpath, logic, component location? RO question Yes NoCan 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 re quire direct entr y to ma jor EOPs?Yes NoDoes 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

• 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 miti gative strate gy of a procedure?No No YesQuestion 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, PA Rs are ____(1)____.

In accordance with EPIP-1, Emergency Plan Classification Logic, ____(2)____

can assume the responsibility for PARs w hen 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 CE CC 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 CE CC 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 CE CC 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 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:

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.

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 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.

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 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 Cl assification.

[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 pl ant events along with communicating significant changes should be performed co ntinuously 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 pers onnel 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 communi cate follow-up information. [2] Reevaluation of significant changes must additionally include a determination of whether Protective Action Recommendat ions (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 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 limits 1 Releases above federally approved limits 1 Release information not known (1Tech Specs/ODCM) Minor releases within federally approved limits 1 Releases above federally approved limits 1 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)

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, C 7, C9, D2, D4, D5, D6, D7, D8, D9 From 26-68 A1, B1, C1, D1, C 7 , D 2, D4, D5 A1, B1, C1, D1, A 3, A4, D2, D3, D4, D5, D6, D7, D8, D9 From 69-110 A1, B1, C1, D1, A 3, D2, D4, D5 A1, B1, C1, D1, A2, A3, A4, A5, A6, A7, D2, D3, D5, D6 From 111-170 A1, B1, C1, D1, A 2, A3, D2, D5 A1, B1, C1, D1, A2, A3, A5, A6, A7, B2, B3, B4, B5, C 2 From 171-230 A1, B1, C1, D1, A 2, A3, B 2, B4, C 2 A1, B1, C1, D1, B 2, B3, B4, B5, C 2, C3, From 231-270 A1, B1, C1, D1, B 2, B4, C 2 A1, B1, C1, D1, B 2, B3, C2, C3, C4, C5, C6, C11 From 271-325 A1, B1, C1, D1, B 2, C2, C4, C5, A1, B1, C1, D1, C2, C4, C5, C6, C7, C8, C9, C10, C11, D 4, D9 From 326-25 A1, B1, C1, D1, C2, C4, C5, C7, C8, D 4 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 limits 1 Minor releases within federally approved limits 1 Releases above federally approved limits 1 Releases above federally approved limits 1 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:

________________________

_______________

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 Em ergency 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 A pproval and PAR development prior to CECC Director ownership for PAR development.

3.1 Precautions

and Limitations CAUTION Unit-2 radiation monitor readings for classi fication 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 Classif ication Level (ECL). 1. For EAL thresholds that specify dur ation 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) 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.

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 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 EMERGENC Y 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)

• 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.

• 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

• 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.

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.

• Administrative processes for temporary modifications.

• Administrative processes for disabling annunciators.

• 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 procedure's overall mitigative strategy or purpose.

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) Page 7 of 16

• 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.

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)

Can the question be answered solely by knowing "systems knowledge", i.e., how the system works, flowpath, logic, component location? RO question Yes NoCan 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 re quire direct entr y to ma jor EOPs?Yes NoDoes 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

• 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 miti gative strate gy of a procedure?No No YesQuestion might not be linked to 10 CFR 55.43(b)(5) for SRO-only NoRO question RO question RO question