ML13214A328

Initial Exam 2013-301 Draft SRO Writen Exam

ML13214A328
Person / Time
Site:	Browns Ferry
Issue date:	08/01/2013
From:	NRC/RGN-II
To:	Tennessee Valley Authority
References
Download: ML13214A328 (167)

Text

QUESTION 76 On April 27 th a series of storms hit the Tennessee Valley, at 0900 Offsite power is lost and the following electrical lineup exists:

• All 500 KV lines are DE-ENERGIZED

• Athens 16 1KV line is DE-ENERGIZED

• Trinity 161KV line is ENERGIZED

• B and D Emergency Diesel Generators failed to start and CANNOT be manually started

• All other equipment has responded as designed At 1000, power has been restored to the Browns Ferry transmission yard via two QUALIFIED 500KV lines (Limestone and Union).

Which ONE of the following is the LATEST date & time Units 1 and 2 are required to be in Mode 4 in accordance with Technical Specification 3.8.1, AC Sources- Operating?

[REFERENCE PROVIDED]

th 27 A. April at2200 th 28 B. April at 2300 th 27 C. April at 2300 th 28 D April at 2200 Answer: B

Level: RO SRO Tier# 1 Group# I Examination Outline Cross-Reference KIA#

295003 AA2.05 Importance Rating 4.2 295003 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER:AA2.05 Whether a partial or complete loss of A.C. power has occurred Explanation: B CORRECT At 0900 3.8.1 Condition J applies (One or more required offsite circuits and two or more unit 1 and 2 diesel generators inoperable). Required Action J. 1 (Enter LCO 3.0.3 Immediately) would require Mode 3 at 2200 on April th 27 and Mode 4 on April 28 th at 2200. One hour later at 1000 when 2 offsite circuits are restored, Condition J no longer applies and LCO 3.0.3 is exited. TS 3.8.1 condition H (Two or more Unit I and 2 diesel generators inoperable) still applies from the original entry time of 0900.

After 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (at 1100), TS 3.8.1 condition I (Required completion time not met) will require Mode 3 at 2300 on April 27th and Mode 4 at 2300 on April 28th.

A Incorrect Plausible because 2200 would be the correct time if 3.0.3 is followed all the way to Mode 4, however the date should be April 28th (the next day).

C Incorrect Plausible because 2300 is the correct time, however the date should be April 28th (the next day).

D Incorrect Plausible because 2200 would be the correct time if 3.0.3 is followed all the way to Mode 4,

the date is correct.

Technical Reference(s): Unit 2 Tech Spec 3.8.1 Proposed references to be provided to applicants during examination: Unit 2 Tech Spec 3.8.1, without bases.

Learning Objective (As available):

Question Source: Bank:

Modified Bank:

New: X Question History: Previous NRC: None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.43(b) 55.43(b) 2 Facility operating limitations in the technical specifications and their bases.

AC Sources Operating

-

3.8.1 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources - Operating LCO 3.8.1 The following AC electrical power sources shall be OPERABLE:

a. Two qualified circuits between the offsite transmission network and the onsite Class I E AC Electrical Power Distribution System;

b. Unit I and 2 diesel generators (DG5) with two divisions of 480 V load shed logic and common accident signal logic OPERABLE; and

c. Unit 3 DG(s) capable of supplying the Unit 3 4.16 kV shutdown board(s) required by LCO 3.8.7, Distribution Systems -

Operating.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS

---

---

NOTE LCO 3.0.4.b is not applicable to DG5.

CONDITION REQUIRED ACTION COMPLETION TIME A. One required offsite A.1 Verify power availability 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> circuit inoperable, from the remaining OPERABLE offsite AND transmission network.

Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter AND (continued)

BFN-U NIT 1 3.8-1 Amendment No. 4T 249 December 1, 2003

AC Sources Operating

-

3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2 Evaluate availability of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> both temporary diesel generators (TDGs).

AND AND Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter 8.3 Declare required 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from feature(s), supported by discovery of the inoperable Unit I and Condition B 2 DG, inoperable when concurrent with the redundant required inoperability of feature(s) are inoperable, redundant required feature(s)

AND 8.4.1 Determine OPERABLE 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Unit I and 2 DG(s) are not inoperable due to common cause failure.

OR 8.4.2 Perform SR 3,8.1.1 for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OPERABLE Unit I and 2 DG(s).

AND (continued)

AC Sources Operating

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3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.5 Restore Unit 1 and 2 DG 7 days from to OPERABLE status, discovery of unavailability of TDG(s)

AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from discovery of Condition B entry

>6 days concurrent with unavailability of TDG(s)

AND 14 days AND 21 days from discovery of failure to meet LCO (continued)

AC Sources Operating

-

3.8.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. One division of 480 V C.1 Restore required division 7 days load shed logic of 480 V load shed logic inoperable, to OPERABLE status.

D. One division of common D.1 Restore required division 7 days accident signal logic of common accident inoperable, signal logic to OPERABLE status.

E. Two required offsite E.1 Declare required 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> from circuits inoperable, feature(s) inoperable discovery of when the redundant Condition E required feature(s) are concurrent with inoperable. inoperability of redundant required feature(s)

AND E.2 Restore one required 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> offsite circuit to OPERABLE status.

(continued)

AC Sources Operating

-

3.8.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

--- ---NOTE NOTE Only applicable when more Enter applicable Conditions and than one 4.16 kV shutdown Required Actions of LCO 3.8.7, board is affected. Distribution Systems -

Operating, when Condition F is entered with no AC power source F. One required offsite to any 4.16 kV shutdown board.

circuit inoperable.

AND F.1 Restore required offsite 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> circuit to OPERABLE One Unit 1 and 2 DG status.

inoperable.

OR F.2 Restore Unit 1 and 2 DG 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to OPERABLE status.

NOTE Applicable when only one 4.16 kV shutdown board is affected.

G. One required offsite G.1 Declare the affected Immediately circuit inoperable. 4.16 kV shutdown board inoperable.

AND One Unit 1 and 2 DG inoperable.

(rnnfini ir1

AC Sources Operating

-

3.8.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME H. Two or more Unit 1 H.1 Restore all but one Unit 1 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 2 DGs and 2 DG to OPERABLE p inoperable, status.

Required Action and 1.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

h. Associated Completion Time of NQ Condition A, B, C, D, E, F, or H not met. 1.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> J. One or more required J.1 Enter LCO 3.0.3. Immediately offsite circuits and two or more Unit I and 2 DGs inoperable.

OR Two required offsite circuits and one or more Unit 1 and 2 DGs inoperable.

OR Two divisions of 480 V load shed logic inoperable.

OR Two divisions of common accident signal logic inoperable.

(continued)

BFN-UNIT I 3.8-6 Amendment No. 234

SRO Only Justification: For the sequence of events presented in the stem of this question, the SRO will be required to assess plant conditions and select the correct Technical Specification Condition that applies to determine when the Required Action Completion Time for MODE 4.

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 /> TSITRM Action?

F Can question be answered solely by knowing the II_____

LCO/TRM information listed above-the-line?

___

Can question be answered solely by knowing the TS Safety Limits?

Does the question involve one or more of the following for TS, TRM, or ODCM?

o Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1) o Application of generic LCO requirements (LCO 10.1 thru 3.01 and SR 4.01 thru 4.0.4) RO-only

• Knowledge of TS bases that is required to analyze TS required actions and terminology No j

I Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only

QUESTION 77 Unit 3 is operating at 100% power when a control air leak occurs on Unit 3.

The Unit 2 to Unit 3 Control Air Crosstie, 2-PCV-032-3901 closes as designed.

Control air pressure continues to drop on Unit 3 and the operators manually scram the reactor.

Not all control rods fully inserted and the APRM DOWNSCALE lights are illuminated.

Which ONE of the following describes the operators plant control actions?

Attempt to insert control rods lAW 3-EOI APPENDIX-_(1)_ and maintain RPV level

_(2)_.

A. (1) iF, MANUAL SCRAM (2) (-)50 to (-)100 inches B. (1)1F,MANUALSCRAM (2) (+)2 to (+)51 inches C. (1) 1E, MANUAL iNSERTION OF CONTROL RODS BY VENTING THE OVER PISTON AREA (2) (-)50 to (-)100 inches D (1) 1E, MANUAL INSERTION OF CONTROL RODS BY VENTING THE OVER PISTON AREA (2) (+)2 to (+)51 inches Answer: I)

Level: RO SRO Tier# I Group# I Examination Outline Cross-Reference 295019 AA2.02 Importance Rating I ?

-- Ability to determine andJor interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUIvIENT AIR: AA2.02 Status of safety related instrument air loads.

Explanation: D CORRECT Inserting Control Rods using appendix IE, MANUAL INSERTION OF

CONTROL RODS BY VENTING THE OVER PISTON AREA, is an available option for inserting control rods. 3-EOI APPENDIX-iF, Manual Scram, resets the scram and drains the scram discharge volume to allow a subsequent scram. Since the CRD Scram discharge volume vent and drain valves (3-FCV 83(83A)(82)(82A)), fail CLOSED on loss of air, draining the SDV is not possible. With APRM DOWNSCALE lights are illuminated, the correct level band is +2 to +51 inches per 3-EOI- 1.

A Incorrect First Part: Incorrect. 3-EOI APPENDIX-iF, Manual Scram, resets the scram and drains the

scram discharge volume to allow a subsequent scram. Since the CRD Scram discharge volume vent and drain valves (3-FCV-85-83(83A)(82)(82A)), fail CLOSED on loss of air, draining the SDV is not possible. Second part: Incorrect. -50 to -100 inches is plausible because this is the level band for rods out and power >5% (APRMs NOT downscale).

B Incorrect First Part: Incorrect. 3-EOI APPENDIX-iF, Manual Scram, resets the scram and drains the

scram discharge volume to allow a subsequent scram. Since the CRD Scram discharge volume vent and drain valves (3-FCV-85-83(83A)(82)(82A)), fail CLOSED on loss of air, draining the SDV is not possible. Second Part: Correct.

C IncorrectFirst Part: Correct. Second part: Incorrect. -50 to -100 inches is plausible because this is the level band for rods out and power >5% (APRMs NOT downscale).

Technical Reference(s): 3-AOI-32-2, 3-EOI APPENDIX- IF, 3-EOI APPENDIX- IE, 3-EOI- 1, 3-EOI C-5 Proposed references to be provided to applicants during examination: None Leaming Objective (As available):

Question Source: Bank:

Modified Bank: X New:

Question History: Previous NRC: Hope Creek 2009 NRC #78 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis: X 10 CFR Part 55 Content: 55.43(b) 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

SRO Only Justification: With the conditions presented in the stem of this question, the SRO will be required to assess the impact of the air system on rod insertion and select 1E, MANUAL INSERTION OF CONTROL RODS BY VENTING THE OVER PISTON AREA as an available method for inserting control rods. Additionally, based on the shutdown status of the reactor (<5% as indicated by APRM downscales), the SRO would need be familiar with the 3-EOI- 1, RPV Control flowchart to direct the correct RPV Level band.

BFN Loss Of Control Air 3-AO[.32-2 Unit3 Rev. 0021 Page 23 of 24 Attachment I (Page 6 of 6)

Expected System Responses 17.0 CORE SPRAY A. 3-FCV-75-57 and 3-FCV-75-58, PSC PUMP SUCTION INBD & OUTBD ISOL VALVES, fail CLOSED on loss of air. ECCS discharge piping pressure must be maintained greater than TRM 3.5.4 limits by the condensate storage and supply system.

18.0 FUEL POOL COOLING A. 3-FCV-78-7, DRAIN VLV TO MAIN CONDENSER, fails CLOSED.

B. FUEL POOL F/DC INFLUENT & EFFLUENT VLVs, 3-FCV-78-19 and 3-FCV-78-26, fail CLOSED.

C FUEL POOL F/D C HOLD PUMP DISCH VLV, 3.-FCV-78-33, fails OPEN.

19.0 CRD A. SCRAM INLET and OUTLET VALVES, 3-FCV-85-39A(B), fail OPEN.

B. EAST & WEST CRD SCRAM DISCH JOL VENT CONT VLVs A & B, 3-FCV-85-83(83A)(82)(82A), fail CLOSED on loss of air.

C. EAST & WEST CRD SCRAM DISCH VOL DRAIN CONT VLVs A & B, 3-FCV-85-37C(D)(E)(F), fail CLOSED on loss of air.

D. CRD SYSTEM FLOW CONTROL VALVES A & B 3-FCV-85-1 1A and 3-FCI-85-1 1 B, fail CLOSED on loss of air. Valves can be manually opened if required.

3-EOI-1, RPV CONTROL RESET ARE.

DEFEAT ARE LOGIC TRIPS IF NECESSARY (APPX 2)

L RCIQ-20 INSERT CONTROL RODS USING ONE OR MORE OF THE FOLLOWING METHODS PLANT CONDrflONS AVAILABLE METHODS APPX DEENERGIZESCRAM SCRAM VALVES SOLENIOIDS FAILED TO OPEN VENT THE 18 SCRAM AIR HEADER

1. RESET SCRAM DEFEAT RPS LOGIC SCRAM VALVES TRIPS IF NECESSARY 4 OPENED BUT 2. DRAIN SDV iF SDV IS FULL 3. RECHARGE ACGU MULATQRS 4, INITIATE RX SCRAM DRIVE CONTROL RODS.

BYPASS RWM IF ID NECESSARY MANUALCONTROL RAISE CRD COOLING IG ROD INSERI1ON WATER HEADER METHODS SCRAM INDIVIDUAL Ic CONTROL RODS A

VENT CONTROL ROD 1E OVER PESTON VOLUMES L

RC/Q-21

RX POWER ASOI/E 5%

OR UNENOWN yssL STOPA1 PREVENT LLINJ INTO RPV ECE?IFR0M RCIO CR0. AND SLC (A?PX 4)

L 05-10 RPVWATER LEVEL DROPS SELOW.5O IN.

C5.29 I

CS-Il IJi6 CONTINUE I I WHILE EXECUTING STEPS C5-15 TO C5-18:

i RE PCWEO ISABOVE5% UM(N01//N 4140 151W WATER LVL IS MOVE - IN.

REIURNIOS1EPCS.0

[::>

L CS-tI

1. 5 154510 EPV NJ MSV CAINE CORE DN/AGE

1. 2 tUltNIEIIAM)VLIIIIEX 1114110

1. 3 ELEVNTEDSIJPPRCFf,IERPRESSMNYORIPRCC

1. 6 1501 OR 15015 5101101-I TEEW MOVE 1411 11 F

L 4,

MAINtAIN 15W/WATER IM EE1AEEN -150 lvi AO 11I IN.

WON THE FOLLOWING ELI 501.115555-151 SITLIRCE OPPX INJ PRESS CNDSMIGLW ES TT1OFSIG ORG SE 184OPSIG ROIl, WITHCST SUCTiON POGGIRLE SC 1245 POlO NW/I, WI IN ITS I 11I..ETIIiN II P1.10210th III 12412 IWIU CR05 GA 410 POlO LPCI 65.60 3S)FSIG SYSTEMS US1ED IS STEP CS-GO CIN.IIiERTTHC&GOl-IOSRPFN PREVIOUSLY PERFORMED L

Figure 2: Screening for SRO-only linked to 10 CFR 85.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, Le, how the system works, flowpath, logic, component location?

I Can the question be answered solely by knowing immediate operator actions? Yes RO question 9

Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs2 Can the question be answered solely by knowing the purpose, overall sequence of events, or overall miti ative strategy of a procedure?

Does the question require one or more of the following?

Q 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 0 Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps Q 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 plart normal, abnormal, and emergency procedures No Question might not be linked to I 10 CFR 5543(b)(5) for SRO-only

Hope Creek 2009 #78 ES-401 Sample Written Examination Form ES-401-5 Question Worksheet Examination Outline Cross-reference: Level RO SRO Tier# 1 Group#

1 KJA# 295019 AA2.02 Importance Rating 3.7 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR : Status of safety-related instrument air system loads Proposed Question: SRO 78 Hope Creek is operating at 100% power when an Instrument Air line in the Turbine Building ruptures. The air compressors are unable to keep up with the loss of air and Instrument Air pressure is lowering.

The operators insert a manual scram.

What will the Reactor Pressure Vessel (RPV) level control and pressure control strategy be for the loss of Instrument Air?

A. JAW EOP-101 RPV Control, SRVs for pressure control, HPCIIRCIC for level control.

B. JAW EOP-101 RPV Control, SRVs for pressure control, Maximize CRDfor level control.

C. JAW A8.ZZ-0000 Reactor SCRAM, Bypass Valves for pressure control, HPCIIRCIC for level control.

D. lAW AB.ZZ-0000 Reactor SCRAM, Bypass valves for pressure control, Maximize CRD for level control.

Proposed Answer: A

QUESTION 78 Unit 3 is in a refueling outage with the final fuel loading in progress.

Division II work is in progress with the following equipment placed under clearance:

• 3-FCV-74-66, RHR SYS II LPCI OUTBD INJECT VALVE

• RHRPump3B C2 RHRSW Pump is tagged out due to a ground on the motor.

The following annunciators are received:

• FUEL POOL SYSTEM ABNORMAL (9-4C, Window 1)

• FLOOD-UP LEVEL ABNORMAL LT-3-55 (9-3E, Window 29)

A report from the refuel floor indicates that the fuel pool level is noticeably LOWERING.

Which ONE of the following describes the MINIMUM operator actions necessary to comply with Technical Specifications?

[REFERENCE FROVIDEDI A. Verify TWO alternate methods of decay heat removal are available AND verify reactor coolant circulation by an alternate method within ONE (1) hour.

B. Immediately suspend movement of fuel assemblies in the RPV, AND start the 3A RHR Pump in shutdown cooling within TWO (2) hours of securing shutdown cooling.

C. Verify ONE alternate method of decay heat removal is available AND EITHER start the 3A RHR Pump in shutdown cooling OR verify reactor coolant circulation by an alternate method within TWO (2) hours.

D. Immediately suspend movement of fuel assemblies in the RPV, verify ONE alternate method of decay heat removal is available within ONE (1) hour, AND verify reactor coolant circulation by an alternate method within ONE (1) hour.

Answer: B

Level: RO SRO Tier# I Group# I Examination Outline Cross-Reference KIA#

295023 G2 .240 Importance Rating I

295023 Refueling Accidents G2.2.40 Ability to apply Technical Specifications for a system Explanation: B CORRECT The fuel pool system abnormal and floodup abnormal together

indicate a low level (per SR-2 the fuel pool abnormal alarm is used to meet the TS requirement to have 22 ft above the flange. This requires immediately suspending fuel movement in the vessel per TS 3.9.6. TS 3.9.8 (RHR Low Water Level) requires two shutdown cooling subsystems (A&C are culTently available) with one in operation but the operating subsystem may be secured for 2 out of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

A Incorrect Plausible since TS 3.9.8 requires this if the 2 required shutdown cooling subsystems are available and does not pay attention to the note that allows circulation to be secured for 2 out of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

C Incorrect Plausible if the operator believes that the non-availability of C2 RHRSW Pump

makes the C subsystem inoperable then an alternate means of decay heat removal would be required.

D Incorrect See A & C Technical Reference(s): TS 3.9.6, 3.9.7, 3.9.9 ARP 9-4C and 9-3E and 3-SR-2 Proposed references to be provided to applicants during examination: TS 3.9.7 page 3.9-14 & TS 3.9.8 page 3.9-18 Learning Objective (As available):

Question Source: Bank:

Modified Bank:

New X Question History: Previous NRC: None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.43(b) 2 Facility operating limitations in the technical specifications and their bases.

___________________

SRO Only Justification: With the conditions presented in the stem of this question, the SRO will be required to assess the impact of the FUEL POOL SYSTEM ABNORMAL (9-4C, Window 1) and FLOOD-UP LEVEL ABNORMAL LT-3-55 (9-3E, Window 29) annunciators on Technical Specifications 3.9.6, RVP Water Level, 3.9.7, RHR- High Water Level, and 3.9.8, RHR- Low Water Level. With knowledge of the surveillance SR-2, the SRO will detennine that fuel movement in the vessel is required to be suspended immediately and that two RHR SDC subsystems (as defined in the Tech Spec Bases) are required to be Operable, and one shall be in operation. Placing an RHR SDC subsystem may be delayed up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> as provided in the NOTE in TS 3.9.8.

Sen.sorlThp Paint FUEL POOL SYSTEM ABNORMAL 3-P1S-078-0011&0016 PumpsAand8LowDisch lOUpsi Preas 3-XA-78-51 3-95-070-0005 Gate Seal or Diywell to S gpm Reactor Well Seal Leakage 3-LS-D7-00025 Fuel Pool Low Lev El 6628 (Page 1 Of 2) 3.-FS-078-0051 Refuefing 8ellows Leakage 5 gpm Sensor 3-PlS-.07a-OQ1I.-0016 :pnl 25-16, El 621, Ccl R-113 S-LINE LocaIion 3-FIS--078-0005 Panel 25-15, El 621, Ccl R-19 S-LINE 3-LS-07S-I3O02B Panel 25-15, El 621, Ccl R-19 S-LINE 3-FS-078-0051 Panel 25-15, El 621, Ccl R-19 S-LINE Prcibable A Pump A and B low discharge pressure (ttakes low discharge pressure on balh Cause: puinçs to cause tfl annunciation to alami

6. Gate seal or driwell to reactor well seal leakage C. Fuel pod lee1 low.

D. Refueling bel[owe leakage.

E Using RWCU Slowdown During Refueling. 3-Ol-9.

F.. Ughting Cabinet 3O6i6reaker 7 Thp (Supply to Panel 25-15 Fuses)

G. 3A Ughting Goar Compartment 201 Trip (Supply to Lighting Cabinet 306).

Automatic Action; Operator A. DISPATCI-1 personnel o Panels 25.15 and 25-16 to determine the ActIon: cause of the alarm. C B. IF Dry Cask acivttiea are being perfamied in the SF5 P. THEN NOTIFY the Cask Supervisor. C C. IF no cause can be determined from Panels 25-IS or 25-1G. THEN VERIFY the following breakers CLOSED:

Lighting Cabinet 3O8lbreaker 7 C 3A 240V LightIng Board Corn pariment 201 C

FUEL POOL SYSTEM ABNORMAL 3-XA-78-51 Window I (Page 2ot2)

Operator Actior: (Continued)

D. [F fuel pool level is low, THEN PERFORM the following:

Cil-IECK weir settiEigs.

AL)D water from Qther urcaa. IF req !Jired. C E IF eaI lealaqe alaim i valid, THEtI PERFORM the following:

1. VERIFY CLOSED 3-DRV-078-055& C

2. VERIFY CLOSED 3-DRV-078-0569. C

3. DETERMINE leak source. C F. IFFPC Pi.mpdlschairge preosure low alarm Is valid, THEN SWAP FPC pumps, REFER TO 3-04-78. C

0. iaci DIRECT personnel to make frequent checks at Panel 2-15 and 2-i6 until this alarm is reset. NPO os,ms C H. [F Fuel Pool CoolIng System failure ias occurred, Ti-lEN REFER TO 3-AOl-78-1. C I. IF thIs alarm is Invalid, THEN REFER TO 0-01-55. C R.eferencaa! GE730E931 Seii 3-45E620-4 3-47E610-78-1 3-47E855-i 87E832-1 3-45E3631-1 1 45N3635-17 FSAR Sections 10.5.4, 10.5.6. 13.6.2 Technical Specfflcaltion Section 3.7.6 Technical Requirements [4anual Sections 3.9.2, 39.3

Serisor/Tno Point Normal Operation 0-200 Inches (÷1-1.5 FLOOD-U P üij .iarrn Ct LEVEL Green sand ABNORMAL 201-500 Inches (÷1-1.5 in.) Alarm In LT-3-55 Yellow Band 3-LT.003O55 RefueF1n Operation 475-484 lncIie (+1-1.5 in.) Afami ln (Page 1 of 1) Yellow 8nd 485489 Inches (÷/-

.5 tin.) Alarm Clear Green Sand 4%-500 Inches (÷1-1.5 in.) Alarm In Red Band Sensor Unit 3 Rx Bldg El. 593 Panel 3-25-SB Cci S-Rh Loca1icn:

PrQ1abie A. Refuel1nQperation Cause:

• Make-Up/Dumpback Mmatoh Loss of Mke.up Dumpbacl

• System misalignment B Normal Operation

• RRRICIHFCl1RCIC lnjecon Automatic A None Action:

OpratQr A Contact refuel floor to verit Rx vessel level and trend C Action: B. If Rx vessel level is high then aØjust make-up!dumpback or secure injection source. [I C. If Fix vessel level is low adjust makelup dumpback, align alternate injection source I3I rered and locate cauee of Iong level. C

References:

347Eg058 3-47E 10-3-2 3-729E89 3-45N3664-2 3-LCI-L-03-055

39 REFUEUNG OPERATIONS 396 Reactor Pressure Vessel (RPV) Water Level LCO 396 RPV water level shall be 22 ft above the top of the RPV flange.

AFPLICABILITY: During movement of irradiated fuel assemblies within the RPV During movement of new fuel assemblies or handling of control rods within the RPV, when irradiated fuel assemblies are seated within the RPV.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A RPV water level not Al Suspend movement of Immediately within limit. fuel assemblies and handling of control rods within the RPV.

RHR-High Water Level 3.9.7 3.9 REFUELING OPERATIONS 3.9.7 Residual Heat Removal (RHR) - High Water Level LCO 3.9.7 One RHR shutdown cooling subsystem shall be OPERABLE and in operation.

NOTE The required RHR shutdown cooling subsystem may not be in operation for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period.

APPLICABILITY: MODE 5 with irradiated fuel in the reactor pressure vessel (RPV) and the water level 22 ft above the top of the RPV flange.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Required RHR shutdown A.1 Verify an alternate 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> cooling subsystem method of decay heat inoperable, removal is available. AND Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter (continued)

BFN-UNIT 3 3.9-14 Amendment No. 212

39 REFUELING OPERATIONS 39.B Residual Heat Removal (RHR) Low Water Level LCO 19.8 Two RHR shutdown cooling subsystems shall be OPERABLE, and one RHR shutdown cooling subsystem shall be in operation.

The required operating shutdown cooling subsystem may not be in operation for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period.

APPLICABILITY: MODE 5 with irradiated fuel in the reactor pressure vessel (RPV) and the water level <22 ft above the top of the RPV flange.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or two required RHR Al Verify an alternate 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> shutdown cooling method of decay heat subsystems inoperable, removal is available for AND each inoperable required RHR shutdown cooling Once per subsystem. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter (continued)

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B Required Action and Si Initiate action to restore Immediately associated Completion secondary containment to lime of Condition A not OPERABLE status met AND B2 Initiate action to restore Immediately two stand by gas treatment subsystems to OPERABLE status AND B3 Initiate action to restore Immediately isolation capability in each required secondary containment penetration fow path not isolated.

(continued)

ACTIONS icontinued CONDITION REQUIRED ACTION COMPLETION TIME C No RHR shutdown Ci Verify reactor coolant 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from coohng subsystem in circulation by an alternate discovery of no cçeration. rnethoci. reactor coolant circulation AND Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter AND C2 Monitor reactor coolant Once per hour temperature.

Attacirnient (Page 15 of 36)

Surveillance Procedure Data Package Modes 4 & 5 TABLE 3.16 REACrORWKTER LEVEL DAY SHIFT V.EEX:

ALrCABLI1Y: Durrn rmvrerrt cr iated E1 assentdte n the RPVn during n erEornewfuei usernblies hig of crdl oriu eithn the PPV, lutien ioarXe &ett aserrtes e sexed eirWothe-RPV. 1ReTo &LSot E.6A) 23onoe Pejrenrets: 3.QjE.1 LOcATION: FleacDx 6ug Btealron 63 boa v(ew InaJa UM1T UO t1lpur (AC(

Fnthy sa Sunday fta&vethetcçoftheR) flange (Note 1 Tuaay Wemoay Thu (1) Wtrea en fue1 d 3ndtheRecorwe( h,e been reraoier. teiia&n tihe cd ow am rFUE(. PCOLSYSTEt6 ASNCRM°J..

3(A-55-4C.Warcow IiHifythaItheIetl ceettThp cftfteRPvfiange.

SURVEILLANCE SR 3.9.6.1 REQUIREMENTS Verification of a minimum water level of 22 ft above the top of the RPV flange ensures that the design basis for the postulated fuel handling accident analysis during refueling operations is met. Water at the required level limits the consequences of damaged fuel rods, which are postulated to result from a fuel handling accident in containment (Ref. 2).

The Frequency of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is based on engineering judgment and is considered adequate in view of the large volume of water and the normal procedural controls on valve positions, which make significant unplanned level changes unlikely.

Figure 1: Screening for SROonly 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 /> TSITRM Action?

1 Can question be answered solely by knowing the LCOITRM information listed above-the-line?

Can question be answered solely by knowing the TS Safety Limits?

Does the question involve one or more of the following for TS, TRM, or ODCM?

0 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 4i4) SRO-only Q Knowledge of TS bases that is required to analyze TS required actions and terminology question No I Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only

QUESTION 79 Unit 1 was in MODE 4 preparing to go to MODE 2. An air leak deeloped-in-theDryweThand resulted in Drywell Pressure of 2.5 psig.

The following indications are observed on the Containment Isolation Status System (CISS) on Panel 1-9-4:

• Groups 1, 3, 4, 5 AND 8 PCIS Logic Success Lights are NOT illuminated

• Groups 2 AND 6 PCIS Logic Success Lights are illuminated The leak has since been isolated AND Drywell Pressure has been restored to 1.3 psig and steady.

Based on these current conditions, which ONE of the following completes the statements?

The Unit Supervisor must direct _(1)_ to be closed.

In accordance with Tech Specs, Unit 1 _(2)_ permitted to change Modes to Mode 2.

A. (1) Reactor Water Cleanup (RWCU) supply and return valves (1 -FCV 1, FCV 69-2 and 1-FCV-69-12)

(2) is NOT B. (1) Reactor Water Cleanup (RWCU) supply and return valves (l-FCV-69-1, FCV 69-2 and 1-FCV-69-12)

(2) is C. (1) Traversing Incore Probe Ball AND Purge valves (2) is D. (1) Traversing Incore Probe Ball AND Purge valves (2) is NOT Answer: C

Level: RO SRO Tier# I Group# I Examination Outline Cross-Reference KIA#

295024 G2.4.21 Importance Rating 4.6 295024 High Drywell Pressure G2.4.2 1 Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

Explanation: C CORRECT High Drywell pressure of 2.45 psig should result in a Group 8 Isolation. CISS

indicates the group failed to isolate as required. With the failure of the automatic function, the crew must take the action to isolate the Traversing Incore Probe Ball and Purge valves. Second Part: TS 3.0.4 states that when an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made 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. TS 3.6.1.3 Condition A must be entered for the TIP Containment Isolation valves but permits continued operation in the MODE for an unlimited period of time.

A Incorrect First Part: Incorrect. Group 3 (RWCU) indicates that a successful isolation did not occur.

However, this group does not isolate on High Drywell Pressure therefore there is no failed automatic action. Plausible in that several Groups do require isolation on high drywell pressure. Second Part:

Incorrect. Plausible in that the plant is going to an applicable mode for PCIVs.

B Incorrect First Part: Incorrect. Group 3 (RWCU) indicates that a successful isolation did not occur.

However, this group does not isolate on High Drywell Pressure therefore there is no failed automatic action. Plausible in that several Groups do require isolation on high drywell pressure. Second Part:

Correct.

D Incorrect First Part: Correct. Second Part: Incorrect. Plausible in that the plant is going to an applicable mode for PCIVs.

Technical Reference(s): U2 TS 3.0.4, TS 3.6.1.3 Proposed references to be provided to applicants during examination: None Learning Objective (As available): OPL1 71.017 V.B.4 Question Source: Bank:

Modified Bank: X New:

Question History: Previous NRC: Browns Ferry 1006 NRC Exam #81 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.43(b) 2 Facility operating limitations in the technical specifications and their bases.

LCO Applicability 3.0 3.0 LCO APPLICABILITY (continued)

LCD 3.0.4 When an LCD is not met, entry into a MODE or other specified condition in the Applicability shall only be made:

a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time;

b. After performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate; exceptions to this Specification are stated in the individual Specifications, or

c. When an allowance is stated in the individual value, parameter, or other Specification.

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.

PCIVs 3.6.1.3 3.6 CONTAINMENT SYSTEMS 3.6.1.3 Primary Containment Isolation Valves (PCIVs)

LCO 3.6.1.3 Each PCIV, except reactor building-to-suppression chamber vacuum breakers, shall be OPERABLE.

APPLICABILITY: MODES 1,2, and 3, When associated instrumentation is required to be OPERABLE per LCO 3.3.6.1, Primary Containment Isolation Instrumentation.

SRO Only Justification: With the conditions presented in the stem of this question, the SRO will be required to recognize and assess the impact of the incomplete Group 8 PCIS isolation. TS 3.6.1.3 Condition A must be entered for the Tll Containment Isolation valves but permits continued operation in the MODE for an unlimited period of time.

Figure 1: Screening for SROorily 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 /> TSITRM Action?

Fn question be answered solely by knowing the I LCOITRM information listed above-the-line?

Can question be answered solely by knowing the TS Safety Limits?

Does the question involve one or more of the following for TS, TRM, or ODCM?

Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)

Application of generic LCO requirements (LCO 30.1 thru 101 and SR40.1 thru4OA) SRO-only

. Knowledge of TS bases that is required to analyze TS question required actions and terminology I I No I Question might not be linked to I 10 CFR 5543(b)(2) for SRO-only

Browns Ferry 1006 NRC Exam #81 HLT 081011006 Written Exam

81. 295024 G2.1.31 Unit 2 is in Mode 4 preparing to go to Mode 2. An air leak in the Drywell result in Drywell Pressure of 2.5 psig. The following indications are observed on the Containment Isolation Status System (CISS) on Panel 2-9-4:

• Groups 2 AND 6 PCIS Logic Success Lights are illuminated

• Groups 1 3, 4, 5 AND 8 PCIS Logic Success Lights are NOT illuminated The leak is subsequently isolated AND Drywell Pressure restored to normal.

Based on these indications, which ONE of the following completes the statements?

The Unit Supervisor must direct _(1)_ valves be closed.

In accordance with Tech Specs, Unit 2 _(2)_ permitted to change Modes to Mode 2.

[REFERENCE PROVIDED]

A. (1) Traversing Incore Probe Ball AND Purge (2) is B. (1) Reactor Water Cleanup (RWCU) suction isolation AND return isolation (2) is C. (1) Traversing lncore Probe Ball AND Purge (2) is NOT D. (1) Reactor Water Cleanup (RWCU) suction isolation AND return isolation (2) is NOT

QUESTION 80 Unit 2 was operating at 100% power when at time 0805 a scram occurred due to a loss of both RPS Bus A and RPS Bus B. NO control rods initially inserted, and, after a manual scram and ART, NOT all controls rods are fully inserted.

At 0817, the Shift Manager, as the Site Emergency Director, made an Emergency Plan event declaration in accordance with EPIP-1, EMERGENCY CLASSIFICATION PROCEDURE.

The following conditions exist:

• Reactor power is UNKNOWN

• Reactor pressure is being maintained 800 to 1000 psig with two (2) SRVs OPEN and a third being manually cycled

• Reactor water level is being maintained -100 to -50 inches using HPCT

• Suppression pool temperature is 136°F and rising Which ONE of the following completes the statement?

The event is a (1) and the State of Alabama must be notified by (2).

[REFERENCE PROVIDED]

A. (1) General Emergency (2) 0832 B. (1) General Emergency (2) 0835 C. (1) Site Area Emergency (2) 0835 D. (1) Site Area Emergency (2) 0832 Answer: D

Level: RO SRO Tier# I Group# 1 Examination Outline Cross-Reference KIA#

295037 G2.4.30 Importance Rating 4.1 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown Knowledge of events relating to system operation/status that must be reported to internal organizations or external agencies, such as the state, the NRC, or the transmission system operator.

Explanation: D CORRECT The automatic scram, manual scram and ARI have failed to make

the reactor subcritical. Power is greater than 5% since greater than 2 SRVs are required to control pressure. Suppression pool temperature has not exceeded HCTL (although it will do so based on present trends) and water level can be maintained above -180 inches, so a general emergency is not required at this time. The time for notification starts when the event classification is declared.

A Incorrect First Part: Incorrect, plausible if the candidate believes that HCTL will be exceeded in the near future and the projection requires a general emergency. Second Part: Incorrect, plausible if the candidate believes that the notification clock starts with the plant conditions reaching the EAL (15 minutes to classify and 15 minutes to notify 30minutes, 0805 +30= 0835).

B Incorrect First Part: Incorrect, plausible if the candidate believes that HCTL will be exceeded

in the near future and the projection requires a general emergency. Second Part: Correct.

C Incorrect First Part: Correct. Second Part: Incorrect, plausible if the candidate believes that the notification clock starts with the plant conditions reaching the EAL( 15 minutes to classify and 15 minutes to notify 3Ominutes, 0805 +30= 0835).

Technical Reference(s): EPIP-1, EPIP-4 Proposed references to be provided to applicants during examination: EPIP- I Learning Objective (As available):

Question Source: Bank:

Modified Bank:

New X Question History: Previous NRC : No Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 5 5.43(b) 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

SRO Only Justification: The SRO will assess plant conditions to determine that the automatic scram, manual scram and ARI have failed to reduce reactor power to < 5% since greater than 2 SRVs are required to control pressure. Additionally the SRO will have to evaluate Suppression pooi temperature and Reactor Pressure to determine HCTL not been exceeded (although it will do so based on present trends). Also water level can be maintained above -180 inches, so a general emergency is not required at this time. From these assessments the SRO will use EPIP-1, Emergency Classification Procedure, to select EAL 1.2-S. Knowledge of the Emergency Notification step in EPIP-4 will be required for the SRO to determine that the time for notification starts when the event classification is declared.

SCRAM FAILURE REACTOR COOLANT ACTIVITY I

I r.2.LUI buId1lL3..uILy j.1..4JJIII a

equivalent 1-131 (Technical Specification Umits) z as detemiinec by chemistty sample. a C

K m

OPERATING CONDITION in ALL z

-4 1.2-A I INOTEI I 1.3-A I I I I Failure of RPS automatic scram functions to bring Reactor coolant activity exceeds 300 iCilgm dose the reactor subcritical equivalent loine-131 as determined by chemistry AND sample.

Manual scram or ARI (automatic or manual> was m siicr.ssfliI OPERATING CONDITION:

OPERATING CONDITION: Model or 2 or 3 Mode 1 or2 1.2-S [ I NOTF I I I I I I 1-allure ot automatic scram manual scram, and 0 ARI to bring the reactor subcritical. -I m

in m

0 m

z OPERATING CONDITION: 0 Model or 2 1.2-G ICURVEI I I us I I I I Failure of automatic scram, manual scram, and ARt. Reactor power is above 3% G)

AND Either of the following conditions exists:

. Suppression Pool temp exceeds HCTL m Referto Curve I .2-G.

. Reactor water level can NOT be restored and maintained at or above -180 inches.

rn OPERA11NG CONDITION:

Mode I or 2

SITF t&RFA FMFRGFNCY FPIP4 iJnitO Pje5o123 3 State of Alabama Nofication NOTE Notification Df the State of Alabama s requied to be completed within 5 minutes from the time of emergency classification dedaration.

SITE AREA EMERGENCY EPIP-4 BFN Rev 0033 UnitO Page 6 of 23 3.4 Notification of the Nuclear Regulatory Commission (NRC)

NOTE Notification of the NRC is required to be completed as soon as possible not to exceed 60 minutes from classification declaration.

Figure 2: Screening for SRO-only linked to 10 CFR 5543(b){6)

(Assessment and selection of procedures)

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

mediate operator actions?

Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?

Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

Does the question require one or more of the following?

D 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 SRO-only

• procedure steps Knowledge of diagnostic steps and decision points in the 1øfs i

t on EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures Knowledge of administrative procedures that specify hierarchy. implementation, and/or coordination of plant normal, abnormal, and emergency procedures No j

Question might not he linked to I 10 CER 55A3(b)(5) for SRO-only

QUESTION 81 All three units are operating at 100% power when a fire is reported in the 4KV Shutdown Board A followed by an explosion.

RHR Pump 2B automatically started and then tripped with no operator action.

Given:

• l-AOI-l00-1, Manual Scram

• 0-SSI-9, Unit 2 Reactor Building Fire 4KV Electrical Board Room 2A

• 3-GOI-100-12A, Unit Shutdown from Power Operation to Cold Shutdown and Reductions in Power During Power Operations Which ONE of the following describes the decision the Shift Manager will direct for each Units emergency/normal shutdown, and governing procedure for each unit?

A. Unit 1 Manual Scram governed by 1-AOI-100-l Unit 2 Manual Scram governed by 0-SSI-9 Unit 3 Unit Shutdown governed by 3-GOI-100-12A B. Unit 1 Manual Scram governed by 1-AOI-100-l Unit 2 Manual Scram governed by 0-SSI-9 Unit 3 Manual Scram governed by 3-AOI-100-l C. Unit 1 Manual Scram governed by 0-SSI-9 Unit 2 Manual Scram governed by 0-SSI-9 Unit 3 Manual Scram governed by 0-SSI-9 D. Unit 1 Manual Scram governed by 0-SSI-9 Unit 2 Manual Scram governed by 0-SSI-9 Unit 3 Manual Scram governed by 3-GO1-100-l2A Answer: C

Level: RO SRO Tier# I Group# 1 Examination Outline Cross-Reference KJA#

600000 AA2. 13 Importance Rating 3.8 Ability to determine and interpret the following as they apply to PLANT FIRE ON SITE AA2. 13 Need for emergency plant shutdown Explanation: C CORRECT Entry conditions of 0-SSI-1 are met for Unit 2 (i.e. any unit greater than atmospheric pressure AND multiple trains of safety related equipment threatened by fire in 4KV Shutdown Board C followed by an explosion). Unit 2 Control Room Operator Actions of 0-SSI-9 directs Units 1 and 3 to perform section 3.0 of 0-SSI-9 to Scram.

A Incorrect Unit 1- incorrect for reasons stated above. Plausible in that this would be the

appropriate procedure to insert a manual scram if the candidate knew that a manual scram was required from 0-SSI-9. Unit 2- Correct. Unit 3- Incorrect. Plausible if the candidate assumes a shutdown is required but doesnt know that 0-SSI-9 requires a scram.

B Incorrect Unit 1- incorrect for reasons stated above. Plausible in that this would be the

appropriate procedure to insert a manual scram if the candidate knew that a manual scram was required from 0-SSI-9. Unit 2- Correct. Unit 3- Incorrect. Plausible in that this would be the appropriate procedure to insert a manual scram if the candidate knew that a manual scram was required from 0-SSI-9.

D Incorrect Unit 1 Correct. Unit 2 Correct. Unit 3- Incorrect. Plausible if the candidate assumes a shutdown is required but doesnt know that 0-SSI-9 requires a scram.

Technical Reference(s): 0-SSI-1, 0-SSI-9 Proposed references to be provided to applicants during examination: None Learning Objective (As available):

Question Source: Bank:

Modified Bank:

New: X Question History: Previous NRC : No Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.43(b) 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

SRO Only Justification: The Shift Manager is the ONLY one who can direct entry into the SSI s. The SRO will need to assess plant conditions , and select the correct procedure with which to proceed. In this case, the SRO will determine the conditions in the stem of the question indicate that a fire in a Unit 2 4kV board room is affecting safety related equipment associated with Unit 2. The SRO will also need to have a working knowledge of O-SSI-l, Safe Shutdown Instructions, well enough to know that this procedure will direct entry into 0-SSI-9, Unit 2 Reactor Building Fire 4KV Electrical Board Room 2A, and direct all units to shutdown via manual scram.

SFN Safe Shutdown Instructions 0-5511.001 Unit 0 Rev. 0014 Page 8 of 109 NOTE The decision to trip the unit(s) and declare an Appendix R tire is left to the judgment of tile Shift Manager, or designee, and must be based on tile magnitude of the fire and potential affect on the safe shutdown equipment necessary to achieve and maintain cold shutdown.

Table 2 provides high pressure systems and available reactor water level instrumentation.

3.0 ENTRY CONDITIONS The sub-instructions are only to be entered when tile following conditions are present.

A. Un[t 1, Unit 2 or Unit 3 reactor is greater than atmospheric pressure, AND B. The Magnitude of the fire has the potential to affect safe shutdown capability as indicated by:

Multiple tailureslspurious actuations of systemsicomponents have occu:rred, OR Erratic or questionable indications on numerous MCR instruments have occurred, OR Multiple trains/channels of safety related equipment are threatened by the fire.

OPL1 71.031 Revision 13 Page 6 of 50 INSTRUCTOR NOTES X. Lesson Body k Purpose The purpose of Safe Shutdown Instructions (SSI) is to Step through 0-provide the actions to ensure safe shutdown of Units 1, 2, SS1-001 and 3 in the event of a major disabling fire. 10 CFR 50 Appendix R transient analyses assume the plant is at full power with normal reactor water level and normal Torus temperature equal to or less than 95°F. The decision to trip the units and declare entry into the SSls is the judgment of the Shift Manager and is based on the magnitude of the fire and potential affect on credited equipment listed in Illustrations I and 2 necessary to achieve and maintain cold shutdown. The SSI analysis is based on a single spurious BEN Unit 2 Reactor Building Fire 4KV O-SSI-S UnitG Electrical Board Room 2A Rev. 0013 Page 6 of 136 INITIALS 2.0 UNIT 2 CONTROL ROOM OPERATOR ACTIONS CAUTION Implementation of Steps 2.0[1] through Step 2.0[21) within 20 minutes is imperative to MINIMIZE the consequences of uncovering the fuel on Unit 2 TBD-2

[1] DIRECT Unit 3 Unit Supervisor to perform Section 3.0 of 0-531-9 to Scram Unit 3, AND PROCEED TO cold shutdown.

[2] DIRECT Unit I Unit Supervisor to perform Section 40 of 0-351-9 to Scram Unit 1, AND PROCEED TO cold shutdown.

NOTE The following instmments are those which have been credited for safe shutdown, and must be referenced when executing manual actions for this fire zone:

2-Ll-3-5A and 2-Pl-3-74A for reactor level and pressure.

TBD-81 2-Ll-64-159A and 2-TI-64-161 for the suppression pool level and temperature.

0Min)

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?

Can the question be answered solely by knowing 1 Yes RO question immediate operator actions? j Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?

Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

Does the question require one or more of the following?

D 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 No I Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only

QUESTION 82 Unit 1 is at 90% power, Unit 2 is in MODE 5 performing a refueling outage, and Unit 3 is at 100% power, when severe weather in the area causes grid instabilities.

You are the Unit Supervisor and the following conditions exist on Unit 1:

• Incoming Mvars are 150 MVAR

• Grid voltage is 505 kV on the 500kV bus

• Grid voltage is 161kV on the 161kV bus

• Grid frequency is fluctuating from 59.97Hz to 60.03Hz The Transmission Operator has notified Browns Ferry that the grid conditions are RED for the 500kV system and YELLOW for the 161kV system.

Which ONE of the following completes the statements?

The required action per 0-AOI I E, Grid Instability, is to (1).

The required Technical Specification action(s) is/are to (2)

A. (1) RAISE reactive power until system voltage returns to 510 KV OR UNTIL Generator Reactive Power reaches +300 MVAR (2) declare ALL offsite circuits Inoperable B. (1) RAISE reactor power by approximately 1%/minute (10 MW(e)/minute) UNTIL systemfrequency returns to 59.98 Hz (2) declare the 161 kV offsite circuits Inoperable, ALL 500 kV offsite circuits remain Operable C. (1) RAISE reactive power until system voltage returns to 510 KY OR UNTIL Generator Reactive Power reaches +300 MVAR (2) declare ALL 500kV offsite circuits Inoperable, the 161 kV offsite circuits remain Operable D. (1) RAISE reactor power by approximately 1%/minute (10 MW(e)/minute) UNTIL system frequency returns to 59.98 Hz (2) declare ALL offsite circuits Inoperable Answer: C

Level: RO SRO Tier# I Group# I Examination Outline Cross-Reference KIA#

700000 AA2.05 Importance Rating I

700000 Generator Voltage and Electric Grid Disturbances AA2.05 Ability to determine andlor interpret the following as they apply to GENERATOR VOLTAGE AND ELECTRICAL GRID DISTURBANCES: Operational status of offsite circuit Explanation: C CORRECT First Part: The required minimum steady-state operating voltage at the BFN

500-ky bus is provided in TRO-TO-SOP-30. 128 applicable Appendix A (distinguished by how many BFN units are tied to the grid). However, the required minimum will be no lower than 515 kV. When a BFN 500-kV bus voltage goes below 515kV and is not corrected within 15 minutes, the TOp shall inform the BFN Generator Operator within 30 minutes from the beginning of the event, that the offsite power source is disqualified (status RED).

A Incorrect Incorrect. The required minimum will be no lower than 515 kV for the 500kV system and 161kV for the 161kV system. Therefore the 161kV system is still qualified. Plausible if the candidate believes that the 161 kV limit is above 161kV just as the 500 kV limit is above 500 kV.

B Incorrect Incorrect. The required minimum will be no lower than 515 kV for the 500kV system and

161kV for the 161kV system. Therefore the 161kV system is still qualified. Plausible if the candidate believes that the 161 kV limit is above 161kV just as the 500 kV limit is above 500 kV.

D Incorrect Incorrect. The 300 Mvar maximum outgoing limit applies to all three units for both 500-ky

and 161 -kVoffsite power source qualification. However, there is no Mvar absorption (incoming) limit for offsite power source qualification. Plausible since 32OMvars stated in the stem would exceed the outgoing limit.

Technical Reference(s): TRO-TO-SOP-30. 128 Proposed references to be provided to applicants during examination: None Learning Objective (As available):

Question Source: Bank:

Modified Bank:

New: X Question History: Previous NRC: None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis: X 10 CFR Part 55 Content: 55.43(b) 2 Facility operating limitations in the technical specifications and their bases.

TRO-TO-SOP-30.128 Browns Ferry Nuclear Plant (BFN) Grid Operating Guide 3.2 Program Elements The grid supplies offsite power to BFN at both the 500-kV aid 161-kV yards. Offsite power must be capable of safely shutting down the plant in the event of a postulated accident and is legally required for plant operation. If the grid cannot assure the ability to meet the plants voltage requirements for a postulated accident, the offsite power source is disqualified, and the plant may be legally compelled to shut clown if offsite power qualification is not restored within a limited amount of time.

Transmission Operator Continually assesses the condition of the BFN offsite power sources and promptly corrects any problems or notifies the plant in accordance with the instructions in this guide and other applicable procedures; informs the BFN Generator Operator of any necessary restrictions on cooling tower load level or alignment; schedules system outages to accbmmodate plant activities and equipment status to the extent practicable; and informs the BFN Generator Operator of real-time and emerging changes to the offsite power supply grid status color code.

3.2 Program Elements The grid supplies offsite power to BFN at both the 500-ky and 161-ky yards. Offsite power must be capable of safely shutting clown the plant in the event of a postulated accident and is legally required for plant operation. If the grid cannot assure the ability to meet the plants voltage requirements for a postulated accident, the offsite power source is disqualified, and the plant may be legally compelled to shut down if offsite power qualification is not restored within a limited amount of time.

Three-way communication is required between BFN and the TOp to ensure that the offsite power qualification status is properly assessed and that any necessary corrective actions are promptly initiated. BFN must inform the TOp of any changes in the voltage regulator or capacitor bank status (remote indication is not available). BFN must also inform the TOp of the request to be considered aligned for 161-ky delayed source, see section 5.3.

A color code system is also used to communicate the status of the offsite power sources, both for current conditions and for postulated grid contingencies. Transmission System Operations, Engineering Analysis group uses a web page to communicate the forecasted status in the Operations Planning time horizon. The TOp verbally informs the plant of status color code changes in real-time.

BFN Cooling Tower loading was incorporated into this study such that the loading is not a factor in determining the offsite power grid status from this guide.

3.2.15 Grid Offsite Power Status Color Codes A color code system is used to communicate the status of each offsite power source, based on the grids ability to support offsite power requirements for current grid conditions as well as following a worst-case transmission contingency. Figure 1 below illustrates the color code decision process.

A RED offsite power grid status means that the source cannot provide qualified offsite power for the given configuration. If the grid offsite power status is RED for actual grid conditions and is not resolved within the 15-minute grace period, the Transmission System Operator must notify the BFN Generator Operator within 30 minutes from the beginning of the event, that offsite power requirements cannot be met for that offsite power source.

3.2.11 Voltage Schedules BEN minimum bus voltage requirements will vary based on system and plant conditions, as determined in the Appendices. There are no upper voltage limits for offsite power qualification.

The required minimum steady-state operating voltage at the BEN 500-ky bus is provided in applicable Appendix A (distinguished by how many BFN units are tied to the grid). The required minimum will be no lower than 515 kV, The official bus voltage for BFN 500 kV source is from bus 1 at BEN.

The required minimum steady-state operating voltage at the BEN 161-ky bus is provided in applicable Appendices B, C, or D (distinguished by immediate or delayed source, BFN capacitor bank and 161 kV yard tied or split status). The required minimum will be no lower than 161-kV.

The official bus voltage for BFN 161 kV source is from bus I at BFN 161 kV yard.

3.2.7 BFN Generator Mvar Limits A 300 Mvar maximum outgoing limit applies to all three units for both 500-ky and 161-ky offsite power source qualification.

There is no Mvar absorption (incoming) limit for offsite power source qualification.

NOTE If the 300 Mvar outgoing Mvar limit is exceeded for a unit and is not corrected within 15 minutes, the Top shall inform the BEN Generator Operator within 30 minutes from the beginning of the event that both offsite power sources are disqualified for the unit that is exceeding the limit (status RED) and notify Engineering Analysis in accordance with TRO-TO-SOP-30.301.

Offsite power qualification is not impaired for the unit(s) whose outgoing Mvars are under the limit.

AC Sources Operating

-

B 3.8.1 BASES LCO Minimum required switchyard voltages are determined by (continued) evaluation of plant accident loading and the associated voltage drops between the transmission network and these loads.

These minimum voltage values are provided to TVAs Transmission Operations for use in system studies to support operation of the transmission network in a manner that will maintain the necessary voltages.

Transmission Operations is required to notify BFN Operations if it is determined that the transmission network may not be able to support accident loading or shutdown operations as required by 10 CFR 50, Appendix A, GDC-17. Any offsite power circuits supplied by that transmission network cannot be credited as a qualified offsite circuit and are inoperable.

SRO Oniy Justification: The SRO will need to assess the current grid conditions to determine why the Transmission Operator is declaring the grid RED for the 500kV system and YELLOW for the 161kV system. He will need to have knowledge of the TRO-TO-SOP-30.128 Browns Ferry Nuclear Plant (BFN)

Grid Operating Guide, limits. From that he will need to know the information in the bases of Technical Specification 3.8.1 regarding the qualification of offsite sources based on this infonnation, and determine that only the 500kV system is NOT Operable for TS 3.8.1.

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

(Tech Specs)

F Can question be answered solely by knowing 1 I hour TSITRM Action?

Can question be answered solely by knowing the LCOiTRM information listed above-the-line?

Can question be answered solely by knowing the TS Safety Limits?

Does the question involve one or more of the following for TS, TRM, or ODCM?

0 Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)

• Application of generic LCD requirements (LCD 301 thru 101 and SR 4.01 thru 4.0.4) I SRO-only 0 Knowledge of TS bases that is required to analyze TS required actions and terminology I question I

No I Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only

QUESTION 83 Unit 2 was operating at 90% when operators inserted a manual reactor scram due to rising secondary containment temperatures and radiation levels as the result of an unisolable Main Steam Line break in the Turbine Building.

The following conditions exist:

• All Rods are fully inserted

• Reactor pressure is 800 psig and slowly lowering

• Reactor Water Level is being maintain +2 to +51 inches with RCIC

• Stack Noble gas (WRGERMS) release rate is 6.1 X 1010 iiCi/sec, and steady for the past 45 minutes

• Field Assessment Teams report the Site Boundary Dose rate is 950 mR/hr gamma, and steady for the past hour Which ONE of the following completes the statements?

The crew (1) required to emergency depressurize the reactor.

The crew is required to declare a (2)

IREFERENCE PROVIDED]

A. (1)is (2) General Emergency B. (1) is (2) Site Area Emergency C. (1)isNOT (2) General Emergency D. (1)isNOT (2) Site Area Emergency ANSWER: A

Level: RO SRO Tier# I Group# 2 Examination Outline Cross-Reference KIA# 295017 AA2.01 Importance Rating 4.2 Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE : Off-site release rate: Plant-Specific Explanation: A CORRECT -When site boundary dose rates reach EPIP General Emergency classification level, and emergency depressurization will help reduce the discharge into primary and secondary containment, 0-EOI-4 requires 2-EOI- 1 be entered and Emergency Depressurization is Required.

B- Incorrect First part is correct. Second part is incorrect because the General Emergency Limits are met

-

for EAL 4.1 -G (Stack Noble gas (WRGERMS) release rate is >5.9 X 1010 j.tCilsec, and steady for the past 15 minutes)

C- Incorrect First part is incorrect because site boundary dose rates have reached EPIP General Emergency

-

classification level and emergency depressurization will help reduce the discharge into primary and secondary containment. Therefore, per 0-EOI-4, 2-EO1- 1 is entered and Emergency Depressurization is Required. Second part is correct.

D- Incorrect First part is incorrect because site boundary dose rates have reached EPIP General Emergency

-

classification level and emergency depressurization will help reduce the discharge into primary and secondary containment. Therefore, per 0-EOI-4, 2-EOI- 1 is entered and Emergency Depressurization is Required. Second part is incorrect because the General Emergency Limits are met for EAL 4.1 -G (Stack Noble gas (WRGERMS) release rate is >5.9 X 1010 tCi/sec, and steady for the l5minutes)

Technical Reference(s): 0-EOI-4, EPIP-1 Proposed references to be provided to applicants during examination: EPIP- 1 Learning Objective (As available):

Question Source: Bank: X Modified Bank:

New:

Question History: Previous NRC: Hatch 2009 question #86 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis : X 10 CFR Part 55 Content: 55.43(b) 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

W.rW-, 1C$4 WHILE EXECISTINC THIS PROCEDURE:

WMIEWRT ntl3C flT&O n.OflOW!13W4O

%1 aMiC1 o1J

,flLaJtO.OflZAIr,J srr,uIn.OrDNG TQr \NTL*.4 L

ri L TABLE S OF FSITE RADOACTIVITV RELEASE CLAS SIFICATI ON UMIT FOR OCNERAL EMEROEI4CY BEFORE TPC Mr1OflEGIACTICO ii tLcE_n:3.3 ,oA GSU2IJS $1AQ NOtE GM FLEtE WP<,HHA1 WW,SOf

_uo1Az$ Nk* ATN CONTINUE L

• cru.. cc WOEIO1W EOCY CLESIFcAION LUll OYThti

• ClOM.OfltWflL ooncO,*.at.ctsAt *NEnOWMAJ. LENCY L E OkNO 0 !IIE yflfl DO4AG&WtO IOU. y,lOynJc ca

.oacoc.Rv

- L

.acnL AT?r1 L

0fl L

I I EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 I BFN I Rev. 0048 I Unit 0 EVENT CLASSIFICATION MATRIX I PAGE 40 OF 205 I NOTES 4.1-U Prior to maidng this emergency classification based upon the WRGERMS indication, assess the release by ether of the following:

I Actual field measurements exceed the limits in table 4.1-U 2.0-SI 4.8.B.1 .a.1 release fraction exceeds 2.0 If neither assessment cai be conducted within 80 minutes then the dedaraton must be made ai the valid WRGERMS reacfing.

4.1-A Prior to making this emergency classification based upon the WRGERMS indication, assess the release by either of the following:

1. Actual field measurements exceed the lkriits in table 4.1-A 2.0-SI 4.8.B.1 .a.1 release fraction exceeds 200 If neither assessment con be conducted witt*i 15 n-iutes then the declaration must be made the valid WRGERMS reading.

4.1-S Prior to making this emergency classification based on the gaseous release rate indication, assess the release by either of the fblb.ing methods:

1. Actual field measurements exceed the limits in table 4.1-S.

2. Prtected or actual dose assessments exceed 100 mrem TEDE or 500 mrem CDE.

If neither assessment can be conducted within 15 minutes then the dedaration must be made based on the valid WRGERMS reading.

4.1-G Prior to making this emergency classification based upon the gaseous release rate indication, assess the release by either of the folbing methods:

1. Actual field measurements exceed the limits in table 4.1-G.

2. Pnected or actual dose assessments exceed 1000 mrem TEDE or 5000 mrem CDE.

If neither assessment con be conducted within 15 minutes then the declaration must be made based on the valid WRGEMS reading.

URVESITARLFS Table 4.1-U RELEASE LIMITS FOR UNUSUAL EVENT TYPE MONITORING METHOD LIMIT DURATION Gaseous Release Rate Stack Noble Gas (WRGERMS) 2.88 X 10 jiCi/sec 1 Hour Gaseous Release Rate 0-SI 4.8.B1.a.1 Release Fraction 2.0 1 Hour Site Boundary Radiation Reading Field Assessment Team 0.10 MREM/HR Gamma 1 Hour Table 4.1-A RELEASE LIMITS FOR I LERT TYPE MI1DIJ 1A1LI.r LIMIT DURATION Gaseous Release Rate Stack Noble Gas (WRGERMS) 2.88 X jiCi/sec 15 Minutes Gaseous Release Rate 0-SI 4.8.B.1.a.1 Release Fraction 200 15 Minutes Site Boundary Radiation Reading Field Assessment Team 10 MREM/HR Gamma 15 Minutes Table 4.1-S RELEASE LIMITS FOR SITE AREA EMERGENCY TYPE MONITORING METHOD LIMIT DURATION Gaseous Release Rate Stack Noble Gas (WRGERMS) 5.8 X 9 i0 iiCi/sec 15 Minutes Site Boundary Radiation Reading Field Assessment Team 100 MREM/HR Gamma 1 Hour Site Boundary lodine-131 Field Assessment Team 3.9 X 10 gCl /cm 1 Hour RELEASE LIMITS FOR GENERAL EMERGENCY TYPE MONITORING METHOD iii DURATION Gaseous Release Rate Stack Noble Gas (WRGERMS) 5.9X 1010 pCi/sec 15 Minutes Site Boundary Radiation Reading Field Assessment Team 1000 MREMJHR Gamma 1 Hour Site Boundary Iodine-131 Field Assessment Team 3.9 X 10 pCI /m

)

3 1 Hour

BFN EMERGENCY CLASSIFICATION PROCEDURE I EPIP-1 I Rev. 0048 Unit 0 EVENT CLASSIFICATION MATRIX I PAGE 41 OF 205 GASEOUS EFFLUENT Description 4.1-U I I NOTE ITABLEI I Gaseous release exceeds ANY limit and duration in Table 4.1-U. z C

0 C

>

I m

OPERATING CONDITION:

ALL 4.1-Al INOTEITABLEI J Gaseous release exceeds ANY limit and duration in Table 4.1-A.

I I

m OPERATING CONDITION:

ALL 4.1-S I I NOTE ITABLEI I 0

EITHER of the following conditions exists:

• Gaseous release exceeds or is expected to exceed ANY limit and duration in Table 4.1-S.

• Dose assessment indicates actual or projected dose consequences above 100 mrem TEDE or 500 mrem thyroid CDE.

z OPERATING CONDITION:

ALL

/ 4.1-G I I NOTE ITABLEI I

-V .. .

C)

EITHER of the following conditions exists: ni

• Gaseous release exceeds or is expected to exceed ANY limit and duration in Table 4.1-S.

• Dose assessment indicates actual or projected dose consequences above 1000 mrem TEDE or 5000 mrem thyroid CDE.

m OPERATING CONDITION ALL

SRO Only Justification: The SRO will assess the given plant conditions and with a knowledge of diagnostic steps and decision points in EOI-4 determine that with an unisolable Main Steam Line break in the Turbine Building and therefore emergency depressurization will help reduce the discharge into primary and secondary containment, O-EOI-4 will reuire 2-EOI-l be entered and Emergency Depressurization is Required. Also, the SRO will use the Stack Noble gas (WRGERMS) release rate and EPIP-l to determine that the General Emergency Limits are met for EAL 4.l-G.

Figure 2: Screening for SRO-only linked to 10 CFR 65.43(b)(5)

(Assessment and selection of procedures)

Can the question be answered solely by knowing systems knowledge, Le, how the system works, flowpath, logic, component location?

Can the question be answered solely by knowing immediate operator actions? Yes RO question

Can the question be answered solely by knowing entry conditions for AO Ps or plant parameters that require direct entry to major EOPs?

4.

Can the question be answered solely by knowing the purpose, overall sequence of events, or iJ4n overall mit!jve strategy of a procedure?

Does the question require one or more of the following?

Q 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 whe:n to implement attachments and appendices, including how to coordinate these items with procedure steps Q Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures

• Knowledgeot administrative procedures that specify hierarchy, implementation, andlor coordination of plant normal, abnormal, and emergency procedures No I Question might not be linked to 10 CER 5543(b)(5) for SRO-only

Hatch 2009 question #86 HLT 4 NRC Exam

86. soim.u.oi ooi Unit 1 was at 80% power when an unisolable Main Steam Line break occurred in the Turbine Building.

The following condthons cmrentiy exit:

o Prompt Offsite Doseresults 1050 mRibrpeak TEDE o Reactor power 411 Rods Full-In o Reactor pressure 900 psig o Reactor Water Level 35 inches Which ONE of the following choices completes both ofthe follong statements?

JAW with 31E0-EOP-014-1, Radinactivity Release Controls EOP flowchart, the crew (l required to emergency depressurize the reactor..

JAW 73EP-EIP-001-0. Tmergency Classification and Initial Actions. the crew (2) -

required to declare a General Emergency A (1)is (2)is B. (l)is (2)isNOT C. (1)isNOT (2)is D. (l)isNOT (2) is NOT At 1000 inRlir the RR chart directs scramming and emergency depressing the reactor. EAL RGI of 73EP-EIP-00 1-0 requires a General Emergency to be declared..

It is Only a Site Area Emergency on the Fission Product Banier Chart..

QUESTION 84 Unit 2 was operating at 100% power when a LOCA occurred.

Current plant conditions are as follows:

• All control rods are inserted

• RPV water level is -135 inches and steady

• RPV pressure is 150 psig and lowering

• Core Spray pump 2A is injecting and is the only makeup source

• Drywell pressure is 32 psig and slowly rising

• Suppression Chamber Pressure is 30 psig and slowly rising

• Suppression Pool Water level is 19.5 feet and slowly rising Based on these conditions, which ONE of the following actions is required to be directed by the Unit Supervisor?

A. Restore Suppression Pool water level to -1 to -6 inches using 2-EOI Appendix- 18 Suppression Pool Water Inventory Removal and Makeup B. Enter EOI C-2, RPV-Emergency Depressurization, and open all 6 ADS SRV s C. Vent the Drywell irrespective of offsite radioactivity release per 2-EOI Appendix 13, Emergency Venting Primary Containment D. Vent the Suppression Chamber irrespective of offsite radioactivity release rate per 2-EOI Appendix 13, Emergency Venting Primary Containment ANSWER: B

Level: RO SRO Tier# I Group# 2 Examination Outline Cross-Reference KIA#

295029 EA2. 03 Importance Rating I

Ability to determine and/or interpret the following as they apply to HIGH SUPPRESSION POOL WATER LEVEL : Drywe 11/containment water level Explanation: B CORRECT 19.5 feet in the Suppression Pool and Suppression Chamber

Pressure at 30 psig is in the action required region of Curve 6, Pressure Suppression Pressure (PSP). EOI-2 step PCIP- 10 questions whether Suppression Chamber pressure can be maintained in the safe area of the curve. The answer is NO and Emergency Depressurization is Required (step PC/P-i 1).

A- Incorrect This is plausible because 19.5 feet in the Suppression Pool and RPV pressure at 150

psig is in the action required region of Curve 4, SRV Tail Pipe Level Limit. The EOI-2 flowchart, SP/L-2 1 questions whether parameters can be restored and maintained in the safe region of the curve. In this case suppression pooi level can be lowered to restore to the safe region. This is not the correct action for C- Incorrect This is plausible if the operator does not identify that Emergency Depressurization

is required. Additionally In the previous revision of the EOI-2 flowchart, step PC/P-i 7 directs venting the DW irrespective of offsite radioactivity release.

D- Incorrect This is plausible because EOI-2 step PC/P-i 3 directs venting the suppression

-

chamber prior to 55psig (step PC/P-12).

Technical Reference(s): 2-EOI-2 Flowchart Proposed references to be provided to applicants during examination: None Learning Objective (As available):

Question Source: Bank:

Modified Bank: X New:

Question History: Previous NRC: Vermont Yankee 2010 NRC #83 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis : X 10 CFR Part 55 Content: 55.43(b) 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

L L

pop-it I VENT the PC to control suppr chnbr rxess below 55 peg (APP 13)

., (ok to exceed rod release rate flits)

L.

PIYP-13

SRO Only Justification: The SRO will assess plant conditions and determine that 19.5 feet in the Suppression Pool and RPV pressure at 150 psig is in the action required region of Curve 6, Pressure Suppression Pressure (PSP). With kinowledge of the decision steps in 2-EOI-2, Primary Containment Control flowchart, the SRO will recognize that Emergency Depressurization is Required and select entry into EOI C-2, RPV-Emergency Depressurization as the procedure with which to proceed.

Figure 2: Screening for SRO-only linked to 10 CFR 5&43(b)(5)

(Assessment and selection of procedures)

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

Can the question be answered solely by knowing immediate operator actions? Yes RO question Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?

d.

Can the question be answered solely by knowing the purpose, overall sequence of events, or question overall mijive strategy of a procedure?

Does the question require one or more of the following?

Q 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 Q 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, andlor coordination of plant normal, abnormal, and emergency procedures No I Question might not be linked to 10 CER 5543(b)(5) for SRO-only

Vermont Yankee 2010 NRC #83 ES-401 Sample Written Examination Form ES-401 -5 L Question Worksheet Examination Outline Cross-reference: Level RO SRO Tier# 1 Group# 2 KJA4t 295029 EA2.03 Importance Rating 3.5 iK&A Statement) EA203- Ability to delorrnine and/or interpret the tollowing as they apply to HIGH SUPRESSK)N POOL.

WATER LEVEL Drywell/contalnmenl water level Proposed Question: SRO 83 Post large break LOCA conditions are as follows:

• AU control rods are inserted

• RPV water level is -35 inches and steady

• RPV pressure is ISO psig and lowering

• Core Spray pump is injecting as designed

• Drywell pressure is 32 psig and rising slow

• Torus Pressure is 30 psig and rising slow

• Torus Water level is 11 .5 feel and rising slow Based on these conditions, which ONE of the following actions is required to be directed by the CRS?

A. Enter the Severe Accident Guideline -i and establish containment flooding B Enter EOP 5, RPV-Emergency Depressunzation and open all SRVs C Vent Primary containment prior to prevent exceeding PCPL-A lAW OE 3107 Appendix HH using the Torus Hardened vent {TVS-86)

0. Vent Primary Containment exceeding Off Site release rates lAW OE 3107, Appendix HH using both trains of SBGT vent through the Stack.

Proposed Answer: B

QUESTION 85 A loss of coolant accident has occurred on Unit 1. The following conditions exist:

• Suppression Pool level is 16 feet

• The 2/0 Analyzer is aligned to the Suppression Chamber and has been in operation for H

13 minutes

• /0 concentrations are as indicated below:

H 2

Which ONE of the following completes the statements?

In accordance with l-EOI-Appendix-19, 2 10 Analyzer Operation, readings from l-XR-76-l 10 H

/O Concentration Recorder (Panel 1-9-54) or from 1-MON-76-1 10, H H

2 /O Analyzer (Panel 1-2 9-55) may only be obtained after (1)

Based on the current 2/O readings and in accordance with 1 -EOI-2, PC/H leg, the crew is H

required to purge the Primary Containment with(2)____

A. (1) 15 minutes (2) 1-EOI-Appendix 14A, N2 MAKEUP TO PRIMARY CONTAINMENT B. (1) 15 minutes (2) 1-EOI-Appendix 14B, CAD OPERATION C. (1) 10 minutes (2) 1-EOI Appendix 14A, N2 MAKEUP TO PRIMARY CONTAINMENT D. (1) 10 minutes (2) 1-EOI-Appendix 14B, CAD OPERATION Answer: C

Level: RO SRO Tier# 1 Group# 2 Examination Outline Cross-Reference K/A#

500000 G2.244 Importance Rating I

500000 High Containment Hydrogen Concentration G2.2.44 Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

Explanation: C CORRECT 1 -EOI Appendix 19 requires the analyzer to be in service for at least 10

minutes before accurate readings can be obtained. Then 1-EO1-2, Primary Containment Control flowchart step PCJH-2 directs control of H2 and 02 using the Nitrogen Makeup System, 1EOI-Appendix 14A.

A Incorrect First part is Incorrect. Plausible if the candidate does not know that 1-EOI-appendix 19 directs

-

10 minutes of operation prior to obtaining accurate concentrations. 15 minutes is a common timeframe for other system operations. Second part is Correct.

B Incorrect First part is Incorrect. Plausible if the candidate does not know that l-EOI-appendix 19 directs

-

10 minutes of operation prior to obtaining accurate concentrations. 15 minutes is a common timeframe for other system operations. Second part is Incorrect, however plausible if the candidate believes that nitrogen addition will be from the CAD system during 1-E0I-2.

D Incorrect First part is Correct. Plausible if the candidate does not know that 1 -E0I-appendix 19 directs

-

10 minutes of operation prior to obtaining accurate concentrations. Second part is Incorrect, however plausible if the candidate believes that nitrogen addition will be from the CAD system during 1-EOI-2.

Technical Reference(s): 1 -EOI-2, 1 -EOI Appendix 19, 1EOI-Appendix 14A Proposed references to be provided to applicants during examination: None Learning Objective (As available):

Question Source: Bank:

Modified Bank:

New: X Question History: Previous NRC: None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis: X 10 CFR Part 55 Content: 55.43(b) 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

_______

I EOI APPENDIX-I9 I I BFN UNIT 1 H2102 ANALYZER OPERATION I - Rev. 2 Page 3 of I

41 T VERWY H2102 ANALYZER SAMPLE PUMP running using 1-XI-76-i1O (Panel 1-9-55).

8. VERIFY LOW FLOW and TROUBLE indicating lights are extinguished on 1-MON-76-11O, H2102 ANALYZER (Panel 1-9-55).

9. WHEN H2102 Analyzer has been aligned and sampling for 10 minutes or greater.

THEN OBTAIN H2 and 02 readings from 1-XR-76-1 10 H202 CONCENTRATION recorder (Panel 1-9-54) or from 1-MON-70-110 H2!02 ANALYZER (Panel 1-9-55)

WI-iI]LE EXECUTING THE FOLLOWING STEPS:

IF TH[N N and O monitoring system is HOTIFY Chem Lab to earn pe For inoperable 2 DWandsupprchmbrHandO Otfste radroavtMty release rate SECURE PC purge reaches 00CM urn its AND OR SECURE PC ver,t NOT reciwed by H2 is 110 longer detected in other EOI steps (24% on confrol room L

PC,H-l VERIFY H2O2 anaIyzr in servie (APPX 1)

L PCH-2 WHEN H is detected in IL IF offiste radioactivity release rate is expected to remain below 00CM limits THEN VENT and PURGE PC as follows:

t. NOTIFY Chem Lab Wsarnple for olfstte radioactivity reease

2. VEMTPC(APPX12}

3. IF PCbevente TI-lEN PURGE PC win, nitrogen m.iteup lAPPX 14A L

PC.H-4 SRO Only Justification: The SRO will assess plant conditions and with a knowledge of 1-EDT Appendix 19, H 0 Analyzer Operation, and 1 -EOI-2, Primary Containment Control flowchart, 2

determine that the analyzer reading is valid and the correct procedure to mitigate! recover would be Nitrogen Makeup System, 1EOI-Appendix 14A.

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

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

(Assessment and selection of procedures)

Can the question be answered solely by <flowing systems knowlecig&, Le, how the system works, tlowpath, logic, component location? i...1:Eestion Can the question he answered ol&y by <flowing 1 immediate operator actions?

] Yes RO question Can the question be answered solely by <flowing entry Donditions for AOPs or plant parameters question

[that require direct entry to major EOPs?

Can the question be answered solely by <flowing the purpose, overall sequence of events, or overa I mitigative strategy of a procedure?

Does the question require one or more of the following?

D 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 S) Knowledge of when to implement attachments and appendices, ncluding how to coordinate these items with procedure steps SRO-only

• Knowledge of diagnostic steps and decision points in the EOPs Ll1-d[ iiiiolve lwiisi&nis to event specilie sub-procedures or emergenc contAngency procedures

• Knowledge of administrative procedures that specify hierarchy. implementation, and/or coordination cf plant normal. abnormal, and energency procedures No I Question might not be linkod to I 10 CFR 5543(b)(5) for SRO-only

QUESTION 86 A startup was in progress on Unit 2 with a nuclear heatup in progress and Reactor Pressure at 550 psig when a LOCA occurred. The reactor scrammed automatically and the following indications now exist:

.-...4rn. fl., nil ..nt ii I.$ø tftAV tYi II AIIUb INI, **tI I

I SPRAY PUMP 28 R POOL SJCT VLV 40L 2.us-7-3oA Which ONE of the following completes the statement?

Core Spray Pump 2B was (I) and the Unit Supervisor will direct (2) to be used for Core Spray Loop II injection.

A. (1) manually stopped (2) 2-01-75, CORE SPRAY SYSTEM B. (1) manually stopped (2) Appendix 6E, INJECTION SUBSYSTEM LINEUP CORE SPRAY SYSTEM II C. (1) automatically tripped (2) 2-01-75, CORE SPRAY SYSTEM D. (1) automatically tripped (2) Appendix 6E, INJECTION SUBSYSTEM LINEUP CORE SPRAY SYSTEM II Answer: B

Level: RO SRO Tier# 2 Group# I Examination Outline Cross-Reference K/A#

209001 G2.1 .31 Importance Rating I

209001 Low Pressure Core Spray System 2.1.31. Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.

Explanation: B CORRECT First Part: Correct. The pump has been manually stopped since the

AMBER light on the vertical section of 9-3 is illuminated. Second Part: Correct. The amber light indicates that the pump has received an accident signal and the EOI are being implemented. The EOIs take precedence over the Ols.

A Incorrect First Part: Correct. The pump has been manually stopped since the AMBER light

on the vertical section of 9-3 is illuminated. Second Part: Incorrect. Plausible since the Ols have a section on auto initiation.

C Incorrect First Part: Incorrect: The AMBER light on the vertical section of 9-3 being

illuminated indicates that the pump handswitch has been placed in STOP with an accident signal present (i.e. manually overridden OFF). Plausible- There has been an auto start of this pump and it is not running now, therefore it is plausible that it tripped automatically. Second Part: Correct.

D Incorrect First Part: Incorrect: The AMBER light on the vertical section of 9-3 being

illuminated indicates that the pump handswitch has been placed in STOP with an accident signal present (i.e. manually overridden OFF). Plausible- There has been an auto start of this pump and it is not running now, therefore it is plausible that it tripped automatically. Second Part: Incorrect. Plausible since the Ols have a section on auto initiation.

Technical Reference(s): 2-EOI- 1, Appendix 6E Proposed references to be provided to applicants during examination: None Learning Objective (As available):

Question Source: Bank:

Modified Bank:

New: X Question History: Previous NRC: None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.43(b) 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

OPLI7I .045 rr

2) If a Cs pump is stopped with an TP-6. 8 initiation signal present, it will not automatically start again until the CS initiation logic is reset. This condition is indicated by an amber light on Panel 9-3 (vertical section). This is accomplished by the stop handswitch energizing K21 which blocks auto start signals until the CS initiation signal is reset dropping out K25.

Note that K21 is also deenergizeci by a loss and restoration of SD boards.

(Loss of offsite power with initiation signal present) such that all pumps will restart when the DGs repower the boards.

SRO Only Justification: The SRO will need to use the pictured control room switches, controls, and indications to assess the plant conditions, and determine that they correctly reflect the desired plant lineup. From that information, the SRO will correctly select the procedure or section of a procedure to mitigate, recover, or with which to proceed, in this case Appendix 6E, INJECTION SUBSYSTEM LINEUP CORE SPRAY SYSTEM II.

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

(Assessment and selection of procedures)

Can the question be answered solly by knowing flowpath, logic, component loca Can the question be answered solely by knowing immediate operator actions? Yes RO question Can the question be answered solely by knowing entry conditions for AOPs or plant parameters questb0n that require direct entry to major EOPs?

Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

Does the question require one or more of the following?

Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps Q 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 No Question might not be linked to I 10 CER 5543(b)(5) for SRO-only

QUESTION 87 Unit 1 is in MODE 3 with all rods fully inserted, following a scram from rated conditions.

Reactor pressure is 940 psig and stable.

While completing post scram actions, annunciator 9-5B window #13, SLC TEMP ABNORMAL, alarms. The AUO reports that local tank temperature is 53° F and the breakers for the heaters are tripped.

SLC storage tank boron concentration is 8.05%.

Given that the temperature will continue to drop while troubleshooting the tripped breakers, which ONE of the following completes the statement?

There are (1) available before reactor coolant temperature must be less than 212 degrees, AND given these plant conditions, the bases for this action is that_(2)__.

[REFERENCE PROVIDED]

A. (1) 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> (2) the standby liquid solution may not meet its design criteria in response to an ATWS B. (1)44hours (2) the standby liquid solution may not meet its design criteria in response to an ATWS C. (1) 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> (2) the standby liquid solution may not meet its design criteria to maintain the proper pH in the suppression pool to control the accident source term following a LOCA D. (1) 44 hours5.092593e-4 days <br />0.0122 hours <br />7.275132e-5 weeks <br />1.6742e-5 months <br /> (2) The standby liquid solution may not meet its design criteria to maintain the proper pH in the suppression pool to control the accident source term following a LOCA ANSWER: D

Level: RO SRO Tier# 2 Group# I Examination Outline Cross-Reference KIA#

211000 A2.05 Importance Rating f 3.4 211000 Standby Liquid Control System A2.05 Ability to (a) predict the impacts of the following on the STANDBY LIQUID CONTROL SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of SBLC tank heaters Explanation: D CORRECT The allowable time is 44 hours5.092593e-4 days <br />0.0122 hours <br />7.275132e-5 weeks <br />1.6742e-5 months <br />. Mode 4 must be achieved due to the LOCA concern.

A- Incorrect The first part, 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> is plausible because candidate needs to know that even though the

plant is currently in mode 3 due to the scram, the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that would have been used to achieve MODE 3 is still available to achieve mode 4 as discussed on page 1.3-4 of the tech spec. bases on completion times (8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of condition B plus only 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of condition C). The second part, the plant is in MODE 3 and SLC is not required to perform its shutdown capability function during MODES 3, 4, or 5.

B- Incorrect First part is correct. Second part, the plant is in MODE 3 and SLC is NOT required to perform its shutdown capability function during MODES 3, 4, or 5. This is plausible because the shutdown capability function is required in modes 1 and 2.

C- Incorrect The first part, 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> is plausible because candidate needs to know that even though the

-

plant is currently in mode 3 due to the scram, the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that would have been used to achieve MODE 3 is still available to achieve mode 4 as discussed on page 1.3-4 of the tech spec. (8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of condition B plus only 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of condition C). Second part, the plant is in MODE 3 and SLC is not required to perform its shutdown capability function during MODES 3, 4, or 5. Both parts are plausible for the reasons stated above.

Technical Reference(s): Unit iTech Spec 3.1.7 and bases Proposed references to be provided to applicants during examination: Unit 1 Tech Spec 3.1.7 Learning Objective (As available):

Question Source: Bank: X Modified Bank:

New:

Question History: Previous NRC: Nile Mile Point U2 2010 NRC #86 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis : X 10 CFR Part 55 Content: 5.43(b) 2 Facility operating limitations in the technical specifications and their bases.

BFN Panel 9-5 1-ARP-9-5B Unit I 1-XA-55-5B Rev. 0018 Page 16 of 42 SensorlTrip Point:

SLC 1-TS-063-0003 Hi-I O5F LO-53.7F suction pipe ABNORMAL 1 -TA-63-3 1-TS-083-0004 Hi-I O5F LO-53.7°F suction pipe (Page 1 of 1)

Sensor l-LPN L-925-0019 Location: Elevation 639 Probable A. Power supply to heaters is open {480V x vent Bd. IA Bkr. SD).

Cause: B. Cleared fuses at 480V Rx vent Sd. 1A.

C. Heater malfunction.

D. Sensor malfunction.

E. Ambient temperature below normal.

F. LC 1O BKR 5 off/tripped.

Automatic None Action:

Operator A. DISPATCH personnel to the standby liquid control tank to Action: DETERMINE if temperature is low in the suction piping or the tank.

B. CHECK Bkr 6D on 480V Rx Vent Bd IA and VERIFY SLC tank heaters energized. C C. CHECK Lighting Cabinet 109 BKR S in ON. C D. REFER TO Technical Specification sect. 3.1.7. C

References:

145E2OS2 147E610631 145E7793 1729E8541, .2

SLC System B 3.1.7 BASES (continued)

AppLicAr h. In MODES I and 2, shutdown capability is required. In MODES 3 and 4, control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied. This provides adequate controls to ensure that the reactor remains subcritical. In MODE 5, only a single control rod can be withdrawn from a core cell containing fuel assemblies. Demonstration of adequate SDM (LCO 3.1 .1, SHUTDOWN MARGIN (SDM)) ensures that the reactor will not become critical. Therefore, the SLC System shutdown capability is not required to be OPERABLE when only a single control rod can be withdrawn.

In MODES 1, 2, and 3, the SLC System must be OPERABLE to ensure offsite doses remain within 10 CFR 50.67, Accident Source Term, limits following a LOCA involving significant fission product releases. The SLC System is used to maintain suppression pool pH above 7 following a LOCA to ensure iodine is retained in the suppression pool water.

SLC System 3.1.7 130 120 110 r____ ACCEPTABLE (PTDV4ed other v1It2nce rcquirncnts are met) oo 90 80 (NOT 70 ACCEPTAB/LE 60 50 40

__

U ju i 30 CONCENTRTI ON (Weight Percent Sodium Pentaborate in Solution)

Figure 3.1.7-1 Sodium Pentaborate Solution Temperature Versus Concentration Requirements

Completion Times 1.3 1.3 Completion Tmes EXAMPLES EXAMPLE 1.3-1 (continued)

The Required Actions of Condition B are to be in MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> AND in MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. A total of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is allowed for reaching MODE 3 and a total of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (not 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />) is allowed for reaching MODE 4 from the time that Condition B was entered. If MODE 3 is reached within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, the time allowed for reaching MODE 4 is the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> because the total time allowed for reaching MODE 4 is 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

If Condition B is entered while in MODE 3, the time allowed for reaching MODE 4 is the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

EXAMPLE 1.3-2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One pump A.1 Restore pump 7 days inoperable, to OPERABLE status.

B. Required 8.1 Be in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Action and MODE 3.

associated Completion NQ Time not met.

B.2 Be in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> MODE 4.

(continued)

BFN-UNIT I 1.3-4 Amendment No. 234

SRO Only Justification: The SRO will assess the annunciator 9-5B window #13, SLC TEMP ABNORMAL, and determine that both trains of Standby Liquid Control are Inoperable per Technical Specification 3.1.7.B. With knowledge of the application of Tech Spec Required Actions (Section 3), the SRO will determine that the Completion Time of 3.1 .7.A is still available from the time the LCO was NOT met. This added to the 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Completion Time in 3.1 .7.C will give a total time allowed to reach MODE 4 of 44 hours5.092593e-4 days <br />0.0122 hours <br />7.275132e-5 weeks <br />1.6742e-5 months <br />. The SRO will need to know the Tech Spec 3.1.7 Bases relative to the current MODE 3 to determine that the reason for the TS Required Action is because standby liquid solution may not meet its design criteria to maintain the proper pH in the suppression pooi following a LOCA.

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

(Tech Specs)

Can question be answered solely by knowing 1 Yes I RO question hour TSITRM Action? I I

Can question be answered solely by knowing the LCOiTRM information listed above-the-line?

Can question be answered solely by knowing the TS Safety Limits?

Does the question involve one or more of the following for TS, TRM, or ODCM?

Q 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 101 thru 10.1 and SR 4.01 thru 4.04) SRO-only Q Knowledge of TS bases that is required to analyze TS required actions and terminology l question I I No I Question might not be linked to 10 CFR 55.43(b)(2) for SRO-only

Nile Mile 2 2010 Question: SRO #86 The reactor is shutdown following a scram from rated conditions. Reactor pressure is 940 psig and stable.

While completing post scram actions annunciator 601711, SLCS TANK 1 TEMPERATURE HIGH/LOW, alarms. The field operator reports that local tank temperature is 69 degrees and that the breakers for both heaters are tripped.

Assuming that temperature continues to drop while troubleshooting the tripped breakers...

(1) How much time is available before reactor coolant temperature must be less than 200 degrees ANL (2) What is the reason for this action given these plant conditions?

A. (1) 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> (2) The standby liquid solution may not meet its design criteria in responding to an ATWS.

8. (1)44hours (2) The standby liquid solution may not meet its design criteria in responding to an A1WS.

C. (1) 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> (2) The standby liquid solution may not meet its design criteria in maintaining the proper pH in the suppression pool following a LOCA.

D. (1)44hours (2) The standby liquid solution may not meet its design criteria in maintaining the proper pH in the suppression pool following a LOCA,

SLC System 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby liquid Control (SLC) System LCO 3.1.7 Two SLC subsystems shall be OPERABLE.

APPLICABILITY: MODES 1,2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SLC subsystem A.1 Restore SLC subsystem 7 days inoperable, to OPERABLE status.

B. Two SLC subsystems B.1 Restore one SLC 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> inoperable, subsystem to OPERABLE status.

C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met.

AND 0.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />

QUESTION 88 Unit 1 is in MODE 2 withdrawing control rods for a startup.

The following conditions exist:

• AlllRMsareonRangel

• lRIvIs G and H are downscale

• All other IRMs are on scale and trending as expected

• SRM D is the highest reading SRM and has reached 1 x 1 cps You are the Unit Supervisor, which ONE of the following actions is required per 1-GOT- 100-1 A, UNIT STARTUP?

A. Bypass SRM D and continue with the startup observing IRMs G and H for proper response.

B. Halt the startup and perform 1-SR-3.3.1.1, IRM Channel Calibration, on IRMs G and H.

C. Declare IRMs G and H Inoperable and continue with the startup.

D. Bypass SRM D and perform l-SR-3.3.l.1, IRM Channel Calibration, on IRMs G and H.

ANSWER: C

Level: RO SRO Tier# 2 Group# I Examination Outline Cross-Reference K/A#

215003 A2.01 Importance Rating 3.2 215003 Intermediate Range Monitor System A2.O 1 Ability to (a) predict tue impacts of the following on the iNTERMEDIATE RANGE MONITOR (IRM) SYSTEM; and (b) based on those predictions, use procedures to orreet,controlorrnjtigate the consequences of those c

7 abnormal conditions or operations: ower supply degrade4 Explanation: C CORRECT Per 1-GOI-100-1A, Unit Startup, and Tech Spec 3.3.1.1 Bases, the overlap between SRMs and IRMs exists when IRM downscale indications have cleared and IRM readings are on scale and trending higher prior to SRMs reaching i0 cps. IRM s G and H are downscale with SRM D is at lXlO cps. IRMs G and H do not meet 1-SR-3.3.1.1.5, and 5

therefore are Inoperable for Tech Spec 3.3.1.1, RPS Instrumentation, function 1 a. However the LCO 3.3.1.1 is still met because there are still 3 required channels per trip system, therefore the startup CAN continue.

A- Incorrect Bypassing SRM D is incorrect. There is nothing in the stem that would indicate

that SRM D is not operating correctly. Plausible because the candidate may think that SRM D indicating 1 x 1 cps is the problem.

B- Incorrect. While IRMs G and H do not meet 1-SR-3.3.1.1.5, and therefore are Inoperable for Tech Spec 3.3.1.1, RPS Instrumentation, function la, the LCO 3.3.1.1 is still met because there are still 3 required channels per trip system, therefore there is no reason to stop and calibrate the IRMs. Doing so at this low power adds additional challenge and could result in a full reactor scram.

D- Incorrect Bypassing SRM D is incorrect. There is nothing in the stem that would indicate

-

that SRM D is not operating correctly. Plausible because the candidate may think that SRM D indicating 1 x cps is the problem. While IRMs G and H do not meet 1-SR-3.3.1.1.5, and therefore are Inoperable for Tech Spec 3.3.1.1, RPS Instrumentation, function 1 a, the LCO 3.3.1.1 is still met because there are still 3 required channels per trip system, therefore there is no reason to stop and calibrate the IRMs. Doing so at this low power adds additional challenge and could result in a full reactor scram.

Technical Reference(s): Unit 1 TS 3.3.1.1 and bases Proposed references to be provided to applicants during examination: None Learning Objective (As available):

Question Source: Bank:

Modified Bank: X New:

Question History: Previous NRC: Brunswick 2007 NRC SRO #88 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis : X 10 CFR Part 55 Content: 55.43(b) 2 Facility operating limitations in the technical specifications and their bases.

B:FN Unit Startup 1-GOI-1004A Unit I Rev. 0032 Page 94 of 194 5.4 Withdrawal of Control Rods while in Mod 2 (continued) 116) VERIFY Reador Engineer records applicable criticality data in 1-SR-3.1 .1.1.

Reactivity Margin Test, Initials Date Time 1171 VERIFY Reactor period greater than 60 seconds.

(R)

Initials Date Time NOTE

1) Completing paper closure of 1-SR-3.3 1.15, SRM and IRM Overlap Verification, is not required prior to performing Step 54[18). However, all AC steps must be VERIFIED

-F, COMPLETED SATISFACTORILY prior to withdrawing SRMs.

2) Tech Spec Bases state that cwetlap between SRMs and IRMs exists wten IRM downscale indicaUons Fiave cleared and IRM readings are on scale and trending higFer prior to SRMs reaching 1 O cps..

1181 VERIFY SRM/IRM overlap by obtaining data and completing 1-SR-S.3.1 .1.5 SRM and IRM Overlap Verification.

(R)

Initials Date Time Reactor Engineer

BFN Unit Startup 1-GOI-100-IA Unit I Rev. 0033 Page 95 of 195 5.4 Withdrawal of Control Rods while in Mode 2 (continued)

NOTES

1) SRMs are fully withdrawn when IRMa are on Range 3 or above and indicating above their downscale trip point.

2) If a shutdown margin test has been performed using a different rod sequence, 1-SR-3.1.3.5(A), Control Rod Coupling Integrity Check, will provide required actions to insert all control rods, establish normal sequence and perform the subsequent start up with re-entry at Step 5.3[13].

102 cps

[19] WITHDRAW SRMs as necessary to maintain them on scale between and cps.

Initials Date Time

[20] MAINTAIN IRMs on scale between approximately 25 and 75 using IRM range switches.

Initials Date Time

[21] ENSURE 1-Sl-4.6.B.1-4, Reactor Coolant Chemistry, has been satisfactorily completed prior to pressurizing Reactor.

(R)

Initials Date Time Chem Shift Supv

[22] WHEN all operable IRMs are on Range 3 or above and all acceptance criteria is met for 1-SR-3.3.1.1.5, SRM and IRM Overlap Verification, THEN WITHDRAW all operable SRMs.

(R)

Initials Date Time

BFN SRM and IRM Overlap Verification 1-SR-3.3.1.1.5 Unit I Rev. 0004 Page 10 of 12 Date 7.0 PROCEDURE STEPS (continued)

NOTES The following table is configured in the same manner as the actual layout of IRM channels while facing Panel 1-9-12.

It is permissible to complete IRM channel status table entries in the following step at a later time if one (1) IRM channel in either or both RPS trip systems is slow clearing its downscale indication.

The following step may be completed when at least three (3) IRM channels per RPS trip system have cleared their downscale indications and are trending higher with all operable SRM channels indicating less than 1 cps.

[8] OBSERVE SRM/IRM channel indications during startup and MARK table below with a checkmark to indicate which IRM channels meet ALL three of the following:

- cleared downscale indication (1-9-5 & 1-9-12 panel IRM downscale lights are not lit),

- indicate 7.5%, and

- are trending higher with all operable SRM channels indicating less than 1 cps.

RPS TRIP SYSTEM A IRM SAT UNSAT IRM SAT UNSAT

--__ C E G RPS TRIP SYSTEM B IRM SAT UNSAT IRM SAT UNSAT B D F H

RPS Instrumentation B 3.31.1 BASES SURVEILLANCE SR 33.11.5 and SR 3.3.1.1.6 REQUIREMENTS (continued) These Surveillances are established to ensure that no gaps in neutron flux indication exist from subcritical to power operation for monitoring core reactivity status.

The overlap between SRMs and IRMs is required to be demonstrated to ensure that reactor power will not be increased into a neutron flux region without adequate indication. This is required prior to withdrawing SRMs from the fully inserted position since indication is being transitioned from the SRMs to

. the IRMs. Overlap between SRMs and IRMs exists when IRM downscale indications have cleared and IRM readings are onscale and trending higher prior to SRMs reaching 1O ops.

SRO Only Justification: The SRO will assess the plant conditions to determine that with IRMs G and H downscale while SRM D is at lXlOcps, IRMs G and H do not meet l-SR-3.3.l.l.5, 5

and therefore are Inoperable for Tech Spec 3.3.1.1, RPS Instrumentation, function 1 a. The SRO will also need to understand the application of generic LCO requirements and recognize that LCO 3.3.1.1 is still met because there are still 3 required channels per trip system, therefore the startup can continue.

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

(Tech Specs)

I Can question be answered solely by knowing 1 flour TSITRM Action?

Can question be wered solelY by knong the Yes LC0TRM information hsted RO quetion Can question be answered solely by knowing the TS Safety Limits?

Does the question involve one or more of the following for TS, TRM, or 00CM?

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 101 thru 10] and SR 401 thftl 404) I SRO-only Knowledge of TS bases that is required to analyze TS question required actions and terminology No I Question might not be linked to I 10 CFR 55.43(b)(2) for SRO-only

RPS Instrumentation 3.3.1.1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1 .1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.1.1-1.

ACTIONS NOTE Separate Condition entry is allowed for each channel.

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 1 of 3)

Reartor Prntctkrn Sytm lnstnimntntion APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTIOJ D.1

1. Intermediate Range Monitors

a. Neutron Flux - High 2 0 SR 3.3.1.1.1 120/125 divisions of full scale i3:1:1:5 SR 3.3.1.1.9 SR 3,3.1.1.14 (a) 5 3 H SR 3.3.1.1.1 120/125 SR 3.3.1.1.4 divisionsoffull SR 3.3.1.1.9 scale SR 3.3.1.1.14

b. Inop 2 3 G SR 3.3.1.1.3 NA SR 3.3.1.1.14 (a) 5 3 H SR 3.3.1.1.4 NA SR 3.3.1.1.14

Brunswick 2007 #88 Proposed Question: Unit One (1) is in Mode 2 withdrawing control rods during a startup.

All IRMs are operable and on Range 1.

lRMCiidT are downscal All other IRMs are reponding as expected.

SRM D is rising and has reached 5 x cps; You are the SCO, which ONE of the following actions is required?

a. In accordance with GP-02 notify the reactor engineer and address T,S. 3.3.11.

b. Bypass SRM D and continue the startup observing IRMs C and F for proper response.

c. Halt the startup and perform I OP-09, Section 8.6 SRM/IRM Response Check on IRMs C and F.

d. Bypass SRM D continue the startup and perform IOP-09, Section 8.6 SRM/IRM Response Check on lRMs C and F.

A

QUESTION 89 Unit 3 is operating at 100% power when a loss of 250V RMOV Board 3B occurs.

Which ONE of the following identifies the required action statement(s), if any, in accordance with Technical Specification 3.5.1, ECCS Operating?

-

[REFERENCE PROVIDED]

A. Action statements G.1 and G.2 are required to be entered B. ONLY action statement E. 1 is required to be entered C. No action statement in LCO 3.5.1 is required to be entered for the supported systems D. Action statement H. 1 is required to be entered Answer: C

Level: RO SRO Tier# 2 Group# I Examination Outline Cross-Reference K/A#

218000 A2.05 Importance Rating 3.6 218000 Automatic Depressurization System Ability to (a) predict the impacts of the following on the AUTOMATIC DEPRESSURJZATION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: A2.05 Loss of A.C. or D.C. power to ADS valves.

Explanation: C CORRECT Per 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.

A Incorrect Plausible because this would be the correct answer if LCO 3.0.6 didnt apply. Two ADS

valves (1-18 and 1-19) are powered from 250V RMOV Bd 3B ONLY. Therefore Required actions G.1 and G.2 would be applicable.

B Incorrect Plausible if the is unaware of LCO 3.0.6 and decides to enter LCO 3.5.1 he/she may believe

that the power supplies are divided up so that no one power supply will cause a loss of power to more than one valve.

D Incorrect Plausible if the candidate is unaware of LCO 3.0.6 and decides to enter LCO 3.5.1 he/she may know two ADS valves (1-18 and 1-19) are powered from 250V RMOV Bd 3B ONLY. Additionally, HPCI is a Division II system and the candidate may believe that 250V RMOV 3B powers Div II. This would lead them to two ADS valves and HPCI Inoperable.

Technical Reference(s): 3-AOl-i-I, Tech Spec 3.5.1 Proposed references to be provided to applicants during examination: Tech Spec 3.5.1, 3.0.6 Learning Objective (As available):

Question Source: Bank:

Modified Bank:

New: X Question History: Previous NRC: None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis: X 10 CFR Part 55 Content: 55.43 (b) (2) Facility operating limitations in the technical specifications and their bases.

LCD 3.0.6 When a supported system LCD 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 LCD ACTIONS are required to be entered. This is an exception to LCD 3.0.2 for the supported system. In this event, an evaluation shall be performed in accordance with Specification 5.5.11, Safety Function Determination Program (SFDP). If a loss of safety function is (continuedi LCD Applicability 3.0 3.0 LCO APPLICABILITY LCD 3.0.6 determined to exist by this program, the appropriate Conditions (continued) and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

When a support systems Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2.

ECCS Operating

-

3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS - Operating LCO 3.5.1 Each ECCS injection/spray subsystem and the Automatic Depress urization System (ADS) function of six safety/relief valves shall be OPERABLE.

APPLICABILITY: MODE 1, MODES 2 and 3, except high pressure coolant injection (HPCI) and ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig.

ACTIONS

---

---

----NOTE LCO 3.0.4.b is not applicable to HPCI.

CONDITION REQUIRED ACTION COMPLETION TIME A. One low pressure ECCS A.1 Restore low pressure 7 days injection/spray subsystem ECCS injection/spray inoperable, subsystem(s) to OPERABLE status.

OR One low pressure coolant injection (LPCI) pump in both LPCI subsystems inoperable.

(continued)

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME G. Two or more ADS valves SI Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> inoperable.

AND OR 52 Reduce reactor steam 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Required Action and dome pressure to associated Completion 150 PSIQ.

Time of Condition C, D, E, or F not met H. Two or more low pressure H.1 Enter LCO 3.0.3. Immediately ECCS injection/spray subsystems inoperable for reasons other than Condition A.

OR HPCI System and one or more ADS valves inoperable.

E One ADS valve El Restore ADS valve to 14 days inoperable. OPERABLE status.

F. One ADS valve Fl Restore ADS valve to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable. OPERABLE status.

AND Condition A entered. F.2 Restore low pressure 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ECCS injection/spray subsystem to OPERABLE status.

(continued)

BFN RliF Vahi Stutk Open 3.AOl..1-1 Unit3 Rev. 0011 Page 27 of 29 Attchnient I (Page I of 3)

Unit 3 SRV Svlenc,id Power BrekeriFue Thble, Panel 25 32 Unit 3 SRV Solenoid Power Breaker Table SERV NORMAL PWR BRKR ALTER\IATE BRKR ALTERNATE BRKR 2SCV RMOV 25CV RMOV BAIL DRD 1i 3A 11B2

=

1-5

• SC TA 2 710 i-1 :fl 1C2

• 1-19 SB 1B2 1-22 3A 1 1 G2 I C1 1-23 3C 181 1-30 3A 1101 1-31 SB BCI 1-34

• SC IGA 2 710 IAI

• B1 3C A 2 710 1-42 SB 3B2 i-irs 313 C2 1-180 3A ID1

• ADS Valve&

All cnrnpnnnFs have hen rmnved from 3-PNI 2

SRO Only Justification: The SRO will assess plant conditions to determine that 2 ADS valves are Inoperable per Technical Specification 3.5.1 .G. The SRO will need to understand the application of generic LCO requirements of LCO 3.0.6 for Supportl Supported systems. LCO 3.0.6 allows entering the Conditions and Required Actions of the SUPPORT system without entering the SUPPORTED system, in this case LCO 3.5.1.

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

(Tech Specs)

I 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 /> TS1TRM Action?

I Can question be answered solely by knowing the LCOITRM information listed above-the-line?

Can question be answered solely by knowing the TS Safety Limits?

Does the question involve one or more of the following for TS, TRM, or ODCM?

0 Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1) 0 Application of generic LCO requirements (LCO 3.01 thru 101 and SR 4.0.1 thru 4.0.4) SRO-only b

0 Knowledge of TS bases that is required to analyze TS required actions and terminology question I____________

No I Question might not be linked to I 10 CFR 55.43(b)(2) for SROonty

QUESTION 90 During a Unit 3 outage, work was performed on Main Steam Relief Valve 3-PCV-1-19.

The work order for 3-PCV-1-19 (ADS valve) requires a Unit Operator to perform a partial surveillance, 3-SR-3 .4.3.2, Main Steam Relief Valves Manual Cycle Test.

Which ONE of the following completes the statements?

This test method (1) an acceptable Post Maintenance Test (PMT) per NPG-SPP-06.3, Pre-/Post-Maintenance Testing.

During tests involving the manipulation of plant equipment, NPG-SPP-06.9.1, Conduct of Testing, requires Operations to be notified (2)

A. (1)is (2) ANY time there is a change of test directors B. (l)isNOT (2) ANY time there is a change of test directors C. (1)is (2) when the test is suspended for more than one shift D. (1)isNOT (2) when the test is suspended for more than one shift Answer: C

Level: RO SRO Tier# 2 Group# I Examination Outline Cross-Reference K/A#

239002 G2 .2.12 Importance Rating 4.1 239002 Relief/Safety Valves 2.2.12 Knowledge of surveillance procedures Explanation: C CORRECT First Part: Per NPG-SPP-06.3, Pre/Post-Maintenance Testing, the

use of a partial surveillance IS an acceptable method of post maintenance testing. Second Part:

Correct: NPG-SPP-06.9. 1, Conduct of Testing, requires Operations to be notified, when the test is suspended for more than one shift.

A Incorrect First Part: Correct. Second Part: Incorrect: Operations Notification is NOT required when changing test directors except for those tests being conducted by Operations personnel located in the Main Control Room.

B Incorrect- First Part: Incorrect. Per NPG-SPP-06.3, Pre/Post-Maintenance Testing, the use of a partial surveillance IS an acceptable method of post maintenance testing. Second Part:

Incorrect: Operations Notification is NOT required when changing test directors except for those tests being conducted by Operations personnel located in the Main Control Room.

D Incorrect First Part: Incorrect. Per NPG-SPP-06.3, Pre/Post-Maintenance Testing, the use of a

-

partial surveillance IS an acceptable method of post maintenance testing. Correct: NPG-SPP 06.9.1, Conduct of Testing, requires Operations to be notified, when the test is suspended for more than one shift.

Technical Reference(s): NPG-SPP-06.3, NPG-SPP-06.9. 1 Proposed references to be provided to applicants during examination: None Learning Objective (As available):

Question Source: Bank:

Modified Bank:

New X Question History: Previous NRC: None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.43(b) 2 Facility operating limitations in the technical specifications and their bases.

NPG-SPP-6.3 Pre/Post Maintenance Testing 3.2.3 PMT Instructions NOTE Surveillance instructions (SI), as specified in this procedure, included other instructions which implement site TS requirements such as surveillance requirements (SR), offsite dose instructions (ODIs), etc.

A. General Requirements

1. An SI may be used as PMT. While an SI may be required forTS operability, supplemental PMT may be required in order to test all components or features either directly or potentially affected by the activity.

2. PMT instructions in a WO shall be specified as follows:

a. Pre-approved plant instructions.

b. Stand-alone portions of pre-approved plant instructions.

c. Specific steps in the body of a WO which satisfy PMT requirements.

d. Specific steps for inspections, visual examinations, data comparisons, or other straight-forward checks.

e. A formal instruction prepared specifically for the WO PMT,

NPG-SPP-06.9.1, Conduct of Testing 3.2.5 Operations Notification Operations shall be notified, during test performance involving the manipulation of plant equipment, of the following items:

A. When startinglstopping the test.

B. When changing test directors except for those tests being conducted by Operations personnel located in the Main Control Room.

C. When suspending the test for more than one shiVt.

D. When continuing a test past an Operations shift change timely notification of the

-

oncoming Operations personnel is required except for those tests being conducted by Operations personnel located in the Main Control Room.

E. When Technical Specification (TS) deficiencies and/or difficulties occur.

F. When changing the expected duration of temporary conditions.

SRO Only Justification: With knowledge of the Pre/Post-Maintenance Testing and the surveillance process, the SRO will recognize a partial surveillance IS an acceptable method of post maintenance testing. From the surveillance test results the SRO will determine that the valve is NOT Operable for Technical Specification 3.5.1, ECCS Operating. 3.5.1 which requires the

-

valve to cycle on a signal (relief function) either from the switch or the logic. However, the SRO will determine that the valve REMAINS Operable for Technical Specification 3.4.3, Safety Relief valves (S/RVs).

Figure 1: Screening for SRO.onIy Unked to 10 CFR 5543(b)(2)

(Tech Specs)

Can question be answered solely by knowing hour TSJTRM Action?

Can question be answered solely by knowing the LCOITRM information listed above-the-line?

Can question be TS Safety Limits?

Does the question involve one or more of the following for TS, TRM, or ODCM?

Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1)

Application of generic LCO requirements (LCO 30.1 thru I

101 and SR 4.0.1 thru 404)

Knowledge of TS bases that is required to analyze TS I SRO-only question required actions and terminology I I No I Question might not be linked to I 10 CFR 55.43(b)(2) for SRO-only

QUESTION 91 An ATWS has occurred on Unit 3. The Unit Supervisor is operating in accordance with 3-C-5, LEVEL AND POWER CONTROL.

The following conditions currently exist:

• RPV water level band is -50 to -100 inches and steady using Feedwater per 3-EOI Appendix -5A, INJECTION SYSTEMS L1NEURCGNDENSATE/FEEDWATER

• RPV Pressure is 900 psig and slowly lowering

• No Boron has been injected

• The Unit Operator ATC is inserting control rods per 3-EOI Appendix-iD, INSERT CONTROL RODS USiNG REACTOR MANUAL CONTROL SYSTEM

• 17 Control Rods are at position 02, all other rods are fully inserted Which ONE of the following describes the actions the Unit Supervisor is required to take, and the basis for that action?

a. Exit 3-EOI-C-5, LEVEL AND POWER CONTROL, enter 3-EOI-i RPV CONTROL, atstep RCIL- 1, and direct RPV Water Level restored to +2 to +51 inches.

The reactor will remain subcritical under all conditions.

b. Remain in 3-EOI-C-5, LEVEL AND POWER CONTROL until all rods are inserted beyond position 02, and maintain current RPV water level band.

The reactor will NOT remain subcritical under all conditions.

c. Exit 3-EOI-C-5, LEVEL AND POWER CONTROL, enter 3-EOI-i RPV CONTROL, at step RC/L-1, and maintain current RPV water level band.

The reactor will NOT remain subcritical under all conditions.

d. Remain in 3-EOI-C-5, LEVEL AND POWER CONTROL until all rods are inserted beyond position 02, and direct RPV Water Level restored to +2 to +51 inches.

The reactor will remain subcritical under all conditions.

ANSWER: A

Level: RO SRO Tier# 2 Group# 2 Examination Outline Cross-Reference KIA#

201003 G2.4. 18 importance Rating 4.0 2001003 Control Rod and Drive Mechanism. Generic 2.4.18 Knowledge of the specific bases for EOPs.

Explanation: A CORRECT The reactor will remain subcritical under all conditions with control

rods in the configuration stated in the stem. The correct action for the Unit Supervisor in this case would be to exit 3-EOI-C-5, LEVEL AND POWER CONTROL, enter 3-EOI-1 RPV CONTROL, at step RCIL- 1. This is stated in the retainment override C5- 1. Once in 3-EOI- 1 RPV CONTROL, the RC/L leg will direct RPV Water Level restored to +2 to +51 inches.

B- Because reactor will remain subcritical under all conditions, it would be incorrect for the Unit Supervisor to remain in 3-EOI-C-5, LEVEL AND POWER CONTROL. However, this is plausible because it would be the correct action for Unit 1. Unit 1 requires all control rods to be inserted to or beyond position 02 to be subcritical under all conditions.

C- The action is correct however the bases is not correct. The reactor will remain subcritical under all conditions with control rods in the configuration stated in the stem. This is plausible because the control rod configuration required to remain shutdown under all conditions is different for different units.

D- Because reactor will remain subcritical under all conditions, it would be incorrect for the Unit Supervisor to remain in 3-EOI-C-5, LEVEL AND POWER CONTROL. However, this is plausible because it would be the correct action for Unit 1. Unit 1 requires all control rods to be inserted to or beyond position 02 to be subcritical under all conditions. The water level will not be restored to +2 to +51 inches in 3-EOI-C-5, LEVEL AND POWER CONTROL.

Technical Reference(s): 3-EOI-C-5, EOIPM 0-V-K Proposed references to be provided to applicants during examination: None Learning Objective (As available):

Question Source: Bank:

Modified Bank:

New: X Question History: Previous NRC: NO Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis : X 10 CFR Part 55 Content: 55.43(b) 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

V Wi-fiLE EXECUTING THIS PROCEDURE:

_I LI RPV water lvi CANNOT be EXIT this procedure and determined ENTER C4, RPV Flooding EXIT this procedure and 40-- The reactor will remain subcritical without boron under all conditions ENTER [01-1, RPV Control at Step ROIL-i PC water lvi CANNOT be maintained below 105 ft STOP inj into the RPV from sources external to the PC NOT required for OR adequate core cooling or shutting Suppr chmbr press CANNOT be down the reactor maintained below 55 psig DW Control Air CROSSTIE CAD to DW becomes unavailable Control Air (APPX 2G)

L C5-1 NOTE The reactor wl I remain subcritical without boron O under all conditions when:

• Any 19 control rods are at position 02 with all other control rods fully inserted OR

• All control rods except one are inserted to or beyond position 00 OR

• Determined by Reactor Engineering EOIPM 0-V-K Flowchart C-5, LEVEL AND POWER CONTROL Bases retainment overide statement directs the operator to transfer RPV water level control actions if it has been determined that the reactor will remain subcritical under all conditions. Engineering calculations have determined that when all control rods are inserted to or beyond position (Maximum Subcritical Banked Withdrawal Position), the reactor will remain subcritical under all conditions.

SRO Only Justification: The SRO will assess the plant conditions and with knowledge of the determine the Notes and Tables in 3-EOI-C-5, LEVEL AND POWER CONTROL, recocognize the requirement to transition from 3-EOI-C-5 to 3-EOI-1 RPV CONTROL, at step RCIL-l, and direct RPV Water Level restored to +2 to +51 inches since 17 Control Rods at position 02, all other rods fully inserted meets the definition of the reactor shutdown under all conditions without boron.

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?

Can the question be answered solely by knowing 1I Yes I I RO question immediate operator actions?

I Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?

Can the question be answered solely by knowing the purpose, overall sequence of events, or iJ.estion overall mitigative strategy of a procedure?

Does the question require one or more of the following?

Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps O 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 No Question might not be linked to I 10 CER 55,43(b)(5) for SRO-only

QUESTION 92 Unit 3 is operating at 100% power when the following occurs:

• Main steam line flow increases rapidly

• All Main Steam Line Isolation Valves close, except MSIV LiNE A iNBOARD, 3-FCV-1-1 4A, which indicates intermediate.

o Main Steam Line Leak Detection High (Panel 9-3D-Window 24) is in alarm and TIS-l 60A reached 240°F and is slowly lowering After conditions have stabilized, which ONE of the following completes the statements?

The Unit Supervisor shall direct (1)

The Emergency Plan classification for this event is (2)

[REFERENCE PROVIDED]

A. (1) Main Steam Line A isolated within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> by verifying CLOSED and de-activating Main Steam Line A Outboard Valve, 3-FCV-1-15 (2) 4.2-U B. (1) Main Steam Line A isolated within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> by verifying CLOSED and de-activating Main Steam Line A Outboard Valve, 3-FCV-1-15 (2) il-S C. (1) Main Steam Line A isolated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by verifying CLOSED and de-activating Main Steam Line A Outboard Valve, 3-FCV-1-15 (2) 4.2-U D. (1) Main Steam Line A isolated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by verifying CLOSED and de-activating Main Steam Line A Outboard Valve, 3-FCV-1-15 (2) 3.1-5 Answer: A

Level: RO SRO Tier# 2 Group# 2 Examination Outline Cross-Reference K/A#

239001 A2. 11 Importance Rating I

239001 Main and Reheat Steam System Ability to (a) predict the impacts of the following on the MAIN AND REI{EAT STEAM SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: A2. 11 Steam line break Explanation: A CORRECT: MSIV LiNE A iNBOARD 3-FCV-l-14 is Inoperable for Technical Specification 3.6.1.3, PCIVs. LCO Condition A and Required Action Al apply. This will direct closing and verifying deactivated the Main Steam Line A Outboard Valve, 3-FCV- 1-15. The Emergency classification for this event is 4.2-S.

B Incorrect First Part: Correct. Second Part: Incorrect. While there is a primary system discharging to Secondary Containment, the area temperature given is below the Maximum Safe Operating Temperature of 3150 F. Plausible if the candidate mistakes the Max Safe Temperature value. 240°F is the Maximum Safe Operating Temperature for other areas of Secondary Containment.

C Incorrect First Part: Incorrect. Plausible since the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> LCO Required Completion time is for other

Primary Containment Penetrations (not Main Steam). Second Part: Correct.

D Incorrect First Part: Incorrect. Plausible since the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> LCO Required Completion time is for other

Primary Containment Penetrations (not Main Steam). Second Part: Incorrect. While there is a primary system discharging to Secondary Containment, the area temperature given is below the Maximum Safe Operating Temperature of 3150 F. Plausible if the candidate mistakes the Max Safe Temperature value.

240°F is the Maximum Safe Operating Temperature for other areas of Secondary Containment.

Technical Reference(s): 3-EOI-1, EPIP-l,Unit 3 TS 3.6.1.3 Proposed references to be provided to applicants during examination: EPIP- 1 ,Unit 3 TS 3.6.1.3 Learning Objective (As available):

Question Source: Bank:

Modified Bank:

New: X Question History: Previous NRC: None Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.43(b) 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

PCIVs 3.6.1.3 3.6 CONTAINMENT SYSTEMS 3.6.1 .3 Primary Containment Isolation Valves (PCIVs)

LCO 3.6.1.3 Each PCIV, except reactor building-to-suppression chamber vacuum breakers, shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, When associated instrumentation is required to be OPERABLE per LCO 3.3.6.1, Primary Containment Isolation Instrumentation.

PCIVs 3.6.1.3 ACTIONS NOTES

1. Penetration flow paths except for 18 and 20 inch purge valve penetration flow paths may be unisolated intermittently under administrative controls.

2. Separate Condition entry is allowed for each penetration flow path.

3. Enter applicable Conditions and Required Actions for systems made inoperable by PCIVs.

4. Enter applicable Conditions and Required Actions of LCO 3.6.1.1, Primary Containment, when PCIV leakage results in exceeding overall containment leakage rate acceptance criteria in MODES 1, 2, and 3.

CONDITION REQUIRED ACTION COMPLETION TIME A. NOTE A.1 Isolate the affected 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> except for Only applicable to penetration flow path by main steam line penetration flow paths use of at least one closed with two PCIVs. and de-activated AND valve, closed manual valve, blind 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for main One or more penetration flange, or check valve steam line flow paths with one PCIV with flow through the inoperable except due to valve secured.

MSIV leakage not within limits.

AND (continued)

BFN EMERQENCY CLASSIFICATION PROCEDURE Unit 0 EVENT CL.ASSIICATION MATRIX PAGE 43 OF 205 MAIN STEAM LINE LIQUID EFFLUENT BREAK Deci ion verition 4.2-U I I 4.3-UI I I I Liquid release rate exceeds 20 times ECL as Main S1em Lir!e biek uu1skJ ueteiiriiried by ilierriistiy iainple z Primarj Containment with solation. C AND Release duration exceeds or will exceed ffl minutes Tn OPERATING CONDITION: OPERATING CONDITION:

Mode or2or3 ALL 4.3-A I I I I Liquid release rate ekceecs 2000 times ECL s determined by chemistry sample AND r

. Tn Release duration exceeds or will exceed 15 minutes. 1 OIhR I IN CcJNLJI I ION:

ALL 4.2S I I I I I I I I ci Unisolable Main Steam Line break outside Primarj Containment.

Tn Tn

G)

Tn OPLRATINC CONDITION:

Modcor2or3 I I I I G

Tn z

n,

I Tn Tn C,

rn z

n

-<

BFN EMERGENCY CLASSIFICATION PROCEDURE ROO4G UnitO EVENT CLASSIFICATION MATRIX PAGE 35 OF 25 SECONDARY CONTAINMENT TEMPERATURE i.scrrpvon I I I I I C

z C

0 C

I m

<

m z

-1 I I I II -

>

m 0

-I 3.i.SI I ITABLEIUSI Cl)

An unisoleble Primary System leak is discharqinq into Secondary Containment m

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BFN Panel 9-3 l.ARP.93D Unit I XA-55-3D Rev. 0026 Page 30 of 43 Sensor/Trip Point:

MAIN STEAM LINE LEAK DETEC11 ON I -TE-1 -60A (1 -TIS-1 -60A) 170°F 1-TA-i -60 1-TE-1-60B through D (1-TR-69-29) 160°F (Page 1 of 2)

Sensor 1-TE-1-SOA Main Steam Tunnel Location: 1-TE-1-60B Main Steam Line Steam Vault Area 21ev 565 I -TE-1-60C Main Steam Line Bypass Vlv Area Elev 586 1-TE-l-60D Main Steam Line Control Vlv Area EIev 617 Probable A. Main Steam RWCU, Feedwater, RCIC, or HPCI discharge (only with HPCI in Cause: service and elevated Suppression Pool water Temp.) line break.

B. TB or RB Coolers out of service C. Sensor malfunction D. Steam Vault Exhaust Booster Fan out of service.

Automatic Impending MSIV Isolation at 189°F area temp.

Action:

Operator ction: NOTE The following steps may be performed in any order or concurrently, as necessary.

A. CHECK the following temperature indicators:

• Main Steam temperature elements on LEAK DETECTION SYSTEM TEMPERATURE Recorder, I -TR-69-29 (Points 13-16) on Panel 1-9-21, 0

• MN STEAM TUNNEL TEMP, 1-TIS-i-60A on 1-9-3, 0

• RWCU Piping in the Main Steam Tunnel temperature indicators, 1 -TIS-69-834A-D, Aux Inst Room I -PN LA-009-0083(84)(85)( 86) or HPTURB mimic on CS. 0 B. CHECK the following flow indications:

• MAIN STEAM LINE FLOWA(B, C, D), 1-Fl-46-1(2, 3, 4) on Panel 1 5.

• RFW FLOW LINE A(B) 1-Fl-3-ThA(B) on Panel 1-9-5. 0

• RFP IA(B, C)flowindicators, 1-Fl-3-20(13, 6) on PneI 1-9-6. 0

SRO Only Justification: The SRO will assess plant conditions and recognize that MSIV LINE A INBOARD 3-FCV-l-14 is Inoperable for Technical Specification 3.6.1.3, PCIVs. He will need to enter LCO Condition A and Required Action Al to close and verify deactivated the Main Steam Line A Outboard Valve, 3-FCV-l-15. Additionally he will have to use the indications given in the stem to classify the event per EPIP-1, and will chose 4.2-S.

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, Le, how the system works, flowpath, logic, component location?

Can the question be answered solely by knowing immediate operator actions?

1 Yes RO question j

Can the question be answered solely by knowing entry conditions for AOPs or plant parameters question that require direct entry to major EOPs?

Can the question be answered solely by knowing the purpose, overall sequence of events, or question overall mitigative strategy of a procedure?

Does the question require one or more of the following?

Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed Knowledge of when to implement attachments and appendices, including how to coordinate these items with 0

procedure steps Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-DL*L3tn procedures or emergency contingency procedures

• Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No I Question might not be linked to I 10 CER 55.43(b)(5) for SRO-only

QUESTION 93 Given the following plant conditions:

• BFN is in the process of discharging the Waste Sample Tank to the river in accordance with an approved Discharge Permit.

• The discharge has been in progress for 90 minutes when Security reports that water is bubbling up from the ground in the vicinity of the Standby Gas Treatment Building.

• 0-RR90- 130, Radwaste Effluent Radiation Monitor, is currently reading the same as the initial background radiation level prior to commencing the discharge.

Which ONE of the following completes the statements below regarding the appropriate action and the basis for this action?

Terminate the discharge and (1)

Enter procedure EPIP-1, EMERGENCY CLASSIFICATION PROCEDURE, in order to determine (2)

A. (1) have Radiation Protection survey the area of the leak (2) the source and isotopic analysis of the leak B. (1) have Radiation Protection survey the area of the leak (2) if Effluent Concentration Limits have been exceeded C. (1) have Chemistry sample the spilled water (2) the source and isotopic analysis of the leak D. (1) have Chemistry sample the spilled water (2) if Effluent Concentration Limits have been exceeded ANSWER: D

Level: RO SRO Tier# 2 Group# 2 Examination Outline Cross-Reference KIA#

68000 A2.01 importance Rating I

Ability to (a) predict the impacts of the following on the RADWASTE ; and (b) basedcnithose predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

System rupture Explanation: D CORRECT: In order to answer this question correctly, the candidate must determine the following: Without references, determine the appropriate notification requirements, procedure entry and basis for a rupture in the radwaste system.

A- A Radiation Protection survey is not required in accordance with EPIP-13, DOSE ASSESSMENT, until directed by EPIP-2, UNUSUAL EVENT; EPIP-3, ALERT; EPIP-4, SITE AREA EMERGENCY; or EPIP-5, GENERAL EMERGENCY.

B- A chemistry sample is appropriate but entrance into EPIP-13, DOSE ASSESSMENT, is not required until an Emergency Classification has been entered. This may or may not occur depending on the results of the chemistry sample.

C- A Radiation Protection is not required in accordance with EPIP- 13 until directed by either EPIP-2, UNUSUAL EVENT; EPIP-3, ALERT; EPIP-4, SITE AREA EMERGENCY; or EPIP-5, GENERAL EMERGENCY. In addition, determination of ECLs (Effluent Concentration Levels) is by chemistry sample, not radiation surveys.

Technical Reference(s): EPIP-1, EPIP-2, EPIP-5 Proposed references to be provided to applicants during examination: None Learning Objective (As available):

Question Source: Bank: X Modified Bank:

New:

Question History: Previous NRC: BFN 0610 #91 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis : X 10 CFR Part 55 Content: 55.43(b) (4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

BFN EMERGENCY CLASSIFICATION PROCEDURE I

I Rev. 0048 Unit 0 EVENT CLASSIFICATION MATRIX PAGE 43 OF 205 MAIN STEAM LINE LIQUID EFFLUENT BREAK Iescription Lescriptiol, 4.2-U I I I I 4.3-U I I I I Liquid release rate exceeds 20 times ECL as Main Steam Line break outside determined by chemistry sample z Primary Containment with isolation.

AND c Release duration exceeds or will exceed 60 minutes.

m OPERATING CONDITION: OPERATING CONDITION:

Model or2or3 ALL I I I I 4.3-Al I I I Liquid release rate exceeds 2000 times ECL as determined by chemistry sample

.

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BFN I EMERGENCY CLASSIFICATION PROCEDURE EPIP-1 Rev. 0046 I

I Uriit0 TECHNICAL BASIS PAGE 144 OF 205 I LIQUID EFFLUENT 4.3-U UNUSUAL EVENT EAL Liquid release rate exceeds 20 times ECL as determined by chemistry sample AND Release duration exceeds or will exceed 60 minutes.

OPERATING CONDrnON: ALL BASiS: Liquid release rates are determined using Surveillance lnstmc-tions which utilize liquid samples rather than nstrumer,t readings for activity determination. Effluent Concentration Limits (ECL) are those annual concentrations given in IOCFR2O Appendix B, Table 2. Column 2. 10 times ECL is equivalent to the instantaneous

h. ODCM limit. Unplanned radioactivity reIee that exceed 20 times ECL (2 times

_ ODCM limit) and continue for 60 minutes or longer represent an uncontrolled situation and potential degradation in the level of safety of the plant. The release should not be averaged over 60 minutes. For example, a release of 40 times ECL for 30 minutes does not meet the requirements of this event classificalion. The 0 minute time period allows sufficient time to isolate any release after exceeding ECL.

Greater than 60 minutes represents inability to isolate or control the release. The Site Emergency Director should declare the event as soon as it is determined that the release duration has or will likely exceed 60 minutes. The Cheniistry Department determines the magnitude of the release by sample procedure tor any release as required by initiatirg procedures (i.e., SI, ARP. A0, EOl). The sample results are reported to the Site Emergency Director as a fraction or multiple of ECL..

Escalation to Alert is based on release in excess of 2000 times ECL for greater than 15 minutes.

REFERENCES:

Reg Guide 1.101 Rev. 3, (NUMARC-AUI example-2)

EDMS L63 010206 800 IOCFR2O NOTES; CURVESITAB LES:

SRO Only Justification: Without references, the SRO will determine that the discharge is the cause of the water is bubbling up from the ground in the vicinity of the Standby Gas Treatment Building. And from that determine that EPIP-l, EMERGENCY CLASSIFICATION PROCEDURE entry will be needed to determine if Effluent Concentration Limits have been exceeded.

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

[10 CFR 55.43(b)(4)j Some examples of SRO exam items for this topic include:

• Process for gaseous/liquid release approvals, i.e., release permits.

QAnalysis 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.

0610 SRO Final Exam

91. Given the following plant condftions:

BFN is in the process of discharging the Waste Sample Tank to the river in accordance with an approved Discharge Permit.

The discharge has been in progress for 90 minutes when Security reports that water is bubbling up from the ground in the vacinity of the SGT Building.

0-RR-90-1 30 (Radwaste Effluent Radiation Monitor) is currently reading the same as the initial background radiation level prior to commencing the discharge.

Which ONE of the following describes the appropriate action and the basis for this action?

Terminate the discharge and (1) . Enter procedure EPIP-1 in order to determine_,

A. (1) have Radcon survey the area of the Teak.

(2) an Offsita Dose Assessment.

B. (1) have Radcon survey the area of the leak.

(2) area dose rates and posting requirements.

C. (1) have Chemistry sample the spiTled water.

(2) the source and isotopic analysis of the leak, D. (1) have Chemistry sample the spilled water.

(2) if the Effluent Concentration Limits have been exceeded.

QUESTION 94 In accordance with O-GOI-100-3C, Fuel Movement Operations During Refueling, which ONE of the following describes where the OFFICIAL cQpyothe Fuel Assembly Transfer Forms (FATF) shall be maintained during fuel handling?

A. The Fuel Handling Supervisors desk B. With the duty Reactor Engineer C. On the refuel platform D. In the Control Room ANSWER: A

Level: RO SRO Tier# 3 Group#

Examination Outline Cross-Reference KIA#

G2.1 .35 Importance Rating G2. 1.35 Knowledge of the fuel-handling responsibilities of SROs.

I Explanation: A CORRECT: In accordance with 0-GOI-100-3C, Fuel Movement Operations During Refueling the copies of the Fuel Assembly Transfer Forms (FATF) shall be maintained at the Fuel

,

Handling Supervisors desk, on the refuel platform, in the Control Room, and at the refuel floor tag board.

The OFFICIAL copy is maintained at the Fuel Handling Supervisors desk.

B- Incorrect. A copy is not required by 0-GOI-100-3C to be kept with the duty Reactor Engineer C- Incorrect. A copy is not required by 0-GOI-100-3C to be kept with the duty Reactor Engineer. Fuel Handling Supervisors desk is missing.

D- Incorrect. A copy is not required by O-GOI-100-3C to be kept with the Work Control SRO. Fuel Handling Supervisors desk is missing.

Technical Reference(s): 0-GOT- 100-3 C Proposed references to be provided to applicants during examination: None Learning Objective (As available):

Question Source: Bank:

Modified Bank:

New: X Question History: Previous NRC: None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis:

10 CFR Part 55 Content: 55.43(b) 6 Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity.

BFN Fuel Movement Operations During O-GOl-100-3C Unit 0 Refueling Rev. 0068 Pane 44 of 122 5.1 Operations During Refueling (continued)

NOTE The copies of the FATFs shall be maintained at their assigned locations.

[2.11 Copies of the FATFs are distributed as follows:

• The official FATE is on the Fuel Handling Supervisors desk.

• The first copy is on the refuel platform.

• The second copy is in the control room.

• The third copy is at the refuel floor tag board (or tag board equivalent).

SRO Only Justification: The SRO will need to know the procedures (in this case O-GOI-100-3C) involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity. In this case specifically related to the details of where the Fuel Assembly Transfer Forms (FATF) are to be maintained during fuel handling.

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

F. Procedures and limitations involved in initial core Ioadinc. alterations in core confiçuration, control rod nrocjramminci. and determination of various internal and external effects on core reactivity. [10 CFR 55.43(b)(6)]

Some examples of SRO exam items for this topic include:

• Evaluating core conditions and emergency classifications based on core conditions.

• Administrative requirements associated with low power physics testing processes.

Q Administrative requirements associated with refueling activities, such as approvals required to amend core loading sheets or administrative controls of potential dilution paths and/or activities.

• Administrative controls associated with the installation of neutron sources.

• Knowledge of TS bases for reactivity controls.

QUESTION 95 Unit 1 is operating at 100% power when a loss of extraction steam to the 1A1 Feedwater Heater is lost.

NO operator actions are taken.

Which ONE of the following describes the plant response and the actions required by the Unit Supervisor?

Reactor power will rise and generator output will (1) slightly.

The Shift Manager must make a report to the NRC within (2).

A. (1)rise (2) 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> B. (1)rise (2) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> C. (1)lower (2) 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> D. (1)lower (2) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Answer: B

Level: RO SRO Tier# 3 Group#

Examination Outline Cross-Reference KIA#

G2. 1.43 Importance Rating I l G2. 1.43 Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc. Z Explanation: B CORRECT -Since it is the #1 heater generator output will go up slightly per 1-AOI 1 A, High Pressure Feedwater Heater String/Extraction Steam Isolation. This will cause the plant to exceed licensed power limits requiring a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> report.

A Incorrect First Part: Correct. Second Part: Incorrect. While 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> reports are required for some items, exceeding a licensed condition requires an 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> report.

C Incorrect First Part: Incorrect. Since it is the #1 heater generator output will go up slightly per

1-AOI-6-1A, High Pressure Feedwater Heater String/Extraction Steam Isolation. Plausible since generator output will lower if any other heater is lost. Second Part: Some discoveries/events are reportable even if they do not actually cause a problem (RPS initiation).

D Incorrect First Part: Incorrect. Since it is the #1 heater generator output will go up slightly per

1 -AOI 1A, High Pressure Feedwater Heater String/Extraction Steam Isolation. Plausible since generator output will lower if any other heater is lost. Second Part: Incorrect. See C Technical Reference(s): 1-AOI-6-1A Proposed references to be provided to applicants during examination: None Learning Objective (As available):

Question Source: Bank:

Modified Bank: X New Question History: Previous NRC: Vermont Yankee 2010 #94 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis: X 10 CFR Part 55 Content: 55.43(b) 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

B. The following symptoms may occur for a closure of an extraction steam isolaon valve:

1. Heater drain cooler flow rises for the next higher heater.

2. Heater drain cooler flow lowers on the affected heater.

3. Feedwater temperature lowers.

4. Reactor power rises.

5. Slight lowering in generator output if affected heater is NOT number one feedwater heater.

BFN High Pressure Feedwater Heater 1AOl-6-1A IJriit I String!Extraetiori Steam Isolation Rev. 0004 Page 4 of 10 2.0 SYMPTOMS (continued)

6. Slight rise in generator output if affected heater is number one feedwater r heater.

NPG Standard Regulatory Reporting Requirements NPG-SPP-03.5 Programs and Rev. 0006 Processes Page 25 af 95 Appendix A (Page 7 of 14)

Reporting of events or Conditions Affecting Licensed Nuclear Power Plants 3.1 Immediate Notification NRC (continued)

-

(i) The results of ensuing evaluations or assessments of plant conditions, (ii) The effectiveness of response or protective measures taken, and (iii) Information related to plant behavior that is not understood.

(2) Maintain an open, continuous communication channel with the NRC Operations Center upon request by the NRC.

3.2 Twenty-Four Hour Notification - NRC Any violation of the requirement contained in specific operating license conditions, shall be reported to NRC in accordance with the license condition.

SRO Only Justification: The SRO will assess the plant conditions to determine that power will rise with a loss of 1A1 Feedwater Heater per 1-A0I-6-IA, and that after the 5% power reduction AOl immediate action, the correct subsequent action the SRO should direct is to refer to 1-01-6 for turbine/heater load restrictions.

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, Le, how the system works, flowpath, logic, component location?

Can the question be answered solely by knowing 1 immediate operator actions?

] Yes RO question Can the question be answered solely by knowing entry conditions for AOPs or plant parameters question that require direct entry to major EOPs?

Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

Does the question require one or more of the following?

Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed Knowledge of when to implement attachments and appendices, including how to coordinate these items with O

procedure steps Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-L*Luestion RO-only procedures or emergency contingency procedures

• Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No I Question might not he linked to 10 CFR 55A3(b)(5) for SRO-only

Vermont Yankee 2010 NRC #94 ES-401 Sample Written Examination Form ES-401 -5 Question Worksheet Examination Outilne Cross-reference: Level RO SAC Tier# 3 Group #

K/A# 2.1.43 Importance Rating 4.3 (K&A Statement> 2.1.43 Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc.

Proposed Question: SRO 94 With a plant startup in progress and operating at 60% RTP, the extraction steam supply from the High Pressure Turbine is lost to the appilcable Heater on the A string.

Mechanical Maintenance has determined the supply will not be restored for another ten hours.

(1) How is the Feedwater Heater string bypassed (2) What is the operational restriction with a Feedwater Heater String bypassed?

A. (1) isolate the heater string lAW OP 2172, Feedwater System (2) lAW OT 3110, Positive ReactIvity Inserhon, reduce reactor power to <23%

RTP B. (1) isolate the heater string lAW OP 2172, Feedwater System (2) lAW OP 0105, Reactor Operations, do not exceed 75% with A Feedwater Heater string isolated.

C. (1) isolate the heater string lAW AP 2170, Condensate System (2> lAW Or 3110, Positive Reactivity Insertion, lAW CT 3110, Positive Reactivity Insertion, reduce reactor power to <23% RTP D. (1) isolate the heater string lAW RP 2170, Condensate System (2) lAW OP 0105, Reactor Operations, do not exceed 75% RTP with A Feedwater Heater string isolated.

Prnnnczcd Anwftr A

QUESTION 96 Unit 2 is in MODE 4. The forced outage-schedule has the unit making a mode change to MODE 2 later this shift. A problem with the RHR 2A pump breaker has just been identified and the repair is estimated to go beyond this shift. The Shift manager has requested a risk assessment be performed per Technical Specification 3.O.4.b to make the mode change to MODE 2.

Which ONE of the following completes the statement?

Per NPG-SPP-09. 11.2, Risk Assessment Methods for Technical Specifications, a MODE Restraint Assessment for Tech Spec 3.0.4(b) can ONLY be used if_______________________

A. there is reasonable likelihood that the RHR 2A will be made Operable within the applicable completion time once the MODE 2 is entered B. a single TS/TRfMaintenance Rule a(4) system/component is impacted C. Operations has identified that RHR 2A is a HIGHER RISK System/Component D. no unusual conditions are present such as impending weather conditions or grid disturbances Answer: A

Level: RO SRO Tier# 3 Group#

Examination Outline Cross-Reference KIA#

G2.2. 18 Importance Rating I

G2.2. 18 Knowledge of the process for managing maintenance activities during shutdown operations, such as risk assessments, work prioritization, etc.

Explanation: A CORRECT: Per NPG-SPP-09. 11.2, Risk Assessment Methods for Technical Specifications, a MODE Restraint Assessment for Tech Spec 3.0.4(b) is used there is reasonable likelihood that the Inoperable equipment will be made Operable within the applicable completion time once the MODE is entered.

B Incorrect Plausible because this is one of the four required criteria listed in NPG-SPP-09. 11.2, Risk Assessment Methods for Technical Specifications, when a risk assessment is NOT required to be performed.

C Incorrect Plausible because this is an SSC category that is specifically prohibited by NPG-SPP-09. 11.2,

Risk Assessment Methods for Technical Specifications, from having a risk assessment performed.

D Incorrect Plausible because this is one of the four required criteria listed in NPG-SPP-09. 11.2, Risk

Assessment Methods for Technical Specifications, when a risk assessment is NOT required to be performed.

Technical Reference(s): NPG-SPP-09. 11.2 Proposed references to be provided to applicants during examination: None Learning Objective (As available):

Question Source: Bank:

Modified Bank:

New: X Question History: Previous NRC: None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.43(b) 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

N PG Standard Risk Assessment Methods for Technical NPG-SPP-09. 11.2 Programs and Specifications Rev. 0000 Processes Page 8 of 19 3.2.1 SR/TSR 3.0.3/4.0.3 - Missed SR Assessment (continued)

C. Qualitative Risk Assessment For SRITSRs related to component or parameters not specifically modeled in the plant PRA or of low risk significance, a qualitative risk assessment may be performed.

3.2.2 LCOITR 3.0.4(b) - MODE Restraint Assessments A. General Requirements

1. TS LCO/TR 3.0.4(b) allows entry into a MODE or specified condition in the applicability with inoperable systems or components; provided a risk assessment is performed and any necessary risk management actions are identified. The risk impact of the MODE change must be assessed and considered, and risk management actions defined as appropriate using the plant programs established to implement Section (a)(4) of the Maintenance Rule (IOCFR5O.65). These programs are defined in NPG-SPP-03.4, Maintenance Rule Performance, Indicator Monitoring, Trending, and Reporting 10CFR5O.65; NPG-SPP-07.1, On-line Work Management; NPG-SPP-07.2.11, Shutdown Risk Management; and NPG-SPP-09.1 1.1, Equipment Out of Service (EOOS) Management.

2. TS LCO/TR 3.0.4 is to be used to go to higher modes of power operation. TS LCO/TR 3.0,4 shall not prevent changes in MODES or other specified conditions in Applicability that are required to comply with actions or part of a shutdown of a unit.

3. This provision should only be used when there is reasonable likelihood that the inoperable equipment will be made Operable within the applicable completion time once the MODE is entered. This provision is intended to be used when unanticipated circumstances occur which would otherwise delay unit startup. It is not intended for routine, intentional use.

4. If a surveillance has not been performed within its specified frequency SR/TSR 3.0.3/4.0.3 provides an allowance to delay declaring the associated LCOITR not met. The delay allows time for the surveillance or a risk assessment to be performed. If the LCO/TR is declared not met then entry into a MODE or specified condition in the Applicability shall be made in accordance with LCO!TR 3.0.4, When an LCO is not met due to Surveillances not having been met, entry into a MODE or other condition in the Applicability shall only be made in accordance with LCOITR 3.0.4.

5. The scope of risk assessments currently performed for IOCFR5O.65 a(4) include equipment as defined in NPG-SPP-03.4, Maintenance Rule Performance, Indicator Monitoring, Trending, and Reporting 10CFR5O,65.

6. The risk assessment performed by Corporate PRA will be in accordance with NEDP-26. All inoperable Technical Specification equipment as well as the risk significant equipment included in the Maintenance Rule (a)(4) scope is included.

The risk assessments performed for TS LCO/TR 3.0.4(b) will use the existing Maintenance Rule (a)(4) scope and NPG-SPP-09.11.1 as a base and explicitly consider, on a case-by-case basis, any additional scope requirements due to existing Technical Specifications Inoperable equipment.

NPG Standard Risk Assessment Methods for Technical NPG-SPP-09.11.2 Programs and Specifications Rev. 0000 Processes Page 10 of 19 3.2.2 LCOITR 3.0.4(b) MODE Restraint Assessments (continued)

-

BFN Only I

2. ;eneric risk assessments have been pertormed the lode change for conditions when only one Tech cal equirements Manual system/component is flop rabk lentified systems for which no MODE chano is how

.e. LCO 3.0.4(b) does not apply) which System MODE to not apply)

Diesel Generators 1, 2, 3 HPCI 1,2 RCIC 1,2

NPG Standard Risk Assessment Methods for Technical NPG-SPP-09.11.2 Programs and Specifications Rev. 0000 Processes Page 11 of 19 3.2.2 LCOITR 3.0.4(b) - MODE Restraint Assessments (continued)

C. Risk Assessment General Requirements

-

1. To determine if the mode change will be allowed refer to Figure 1 for a flow chart giving the general approach to TS LCO/TR 3.0.4(b) risk evaluation.

NOTE All modes changes made using the TS LCOITR 3.0.4(b) provision must be approved by the Plant Manager or Designee and documented on Attachment 1.

2. Mode changes are allowed without a risk assessment provided that all four of the following items are true:

a. Only a single TSITR/Maintenance Rule a(4) system/component impacted.

b. The system/component is not a high risk system.

c. There is a high probability that the work involved can be completed within the expected completion time.

d. No unusual conditions are present such as impending weather conditions or grid disturbances.

SRO Only Justification: The SRO must have knowledge of NPG-SPP-9. 11.2 Risk Assessment Methods for Technical Specifications.

NPG Standard Risk Assessment Methods for Technical NPG-SPP-09.11.2 Programs and Specifications Rev. 0000 Processes Page 14 of 19 Figure 1 (Page 1 of 1)

Flow Chart for Risk Assessment Operations to determine lOperationato determine Mode LCOITR(s) which will L_Jacceptabilityof 0

_ JChange is become applicable but not met 9LCO!TR(s) 3.0.4(b) E 9NOT applicotion allowed.

Yes Maintenance to 1 determine reasonable Yes Operations to determine lilt Is a System/Component identified as higher risk in es Section 3.2.2 (B)?

No Multiple Systems! Single TSITR(s) or (a)(4)

Components scope System!

within TSITR(s) or (a)(4) Component scope impacted impacted Yea Arethere any unusual conditions e.g.

I - -inpendirrg weather condition

[ognd clistuthances NEDP26evakideonto NO

_No_aJ Mode Change Yes Is allowed Identify and establish risk management actions as Mode Change is I appropnate L

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, Le, how the system works, flowpath, logic, component location?

Can the question be answered solely by knowing 1 Yes RO question s

immediate operator actions?

Can the question be answered solely by knowing entry conditions for AOPs or plant parameters J+jestion that require direct entry to major EOPs?

Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

Does the question require one or more of the following?

• Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with RO-only

• procedure steps Knowledge of diagnostic steps and decision points in the DLution EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No I Question might not be linked to I 10 CER 55.43(b)(5) for SRO-only

QUESTION 97 In accordance with EOI Flowchart C-i, ALTERNATE LEVEL CONTROL, which ONE of the following completes the statement?

The Minimum Zero Injection RPV Water Level, MZIRWL, value is (1 )_ inches for (2)_.

A. (1) (-)200 (2) all three Units B. (l)(-)195 (2) Units 1 and 2 ONLY C. (1)(-)195 (2) Units 2 and 3 ONLY D. (i)(-)200 (2) Unit 2 ONLY ANSWER: C

Level: RD SRO Tier# 3 Group#

Examination Outline Cross-Reference KIA#

G2.2.3 importance Rating G2.2.3 Knowledge of the design, procedural, and operational differences between units.

I Explanation: C CORRECT: -195 inches is the Minimum Zero Injection Reactor Water Level for Units 2 and 3.

A- Incorrect. First Part: Correct. Second Part: Incorrect, -200 inches is the Minimum Zero Injection Reactor Water Level for Unit 1, NOT all three Units. Plausible in that the candidate may not know there is a difference between units.

B- Incorrect. Plausible since -195 inches is the Minimum Zero Injection Reactor Water for Units 2 and 3, NOT Units 1 and 2.

D- Incorrect. Plausible since -200 inches is the Minimum Zero Injection Reactor Water for Units 1, NOT Unit 2.

Technical Reference(s): 1 (2,3)-EOI Flowchart C-4 Proposed references to be provided to applicants during examination: None Learning Objective (As available):

Question Source: Bank:

Modified Bank:

New: X Question History: Previous NRC: No Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis:

10 CFR Part 55 Content: 55.43(b) 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

OPL171.205 EOI CONTINGENCIES

18. Step 01-18 This decision step is reached from step 01-17 or from step C1-21 (third override).

RPV water level at this point is somewhere between -180 and MZIRWL. UNIT DIFFERENCE MZIRWL is:

Unit 1 -200 Unit 2 -195 Unit 3 -195

a. If this decision step was reached from the steam cooling override (Cl-21), and water level cannot be restored and maintained above -180 then emergency depressurization is required at this point because:

1) The calculation for the MZIRWL assumes that there is no MZIRWL, Minimum subcooling at the core inlet. Since any injection would invalidate Zero Injection this assumption, adequate core cooling cannot be assured if RPV Reactor Water water level is below -180 and water is being injected into the Level is defined in RPV. OPL17I.201, Introduction to 601s

UNIT 1 V

STEAM COOLNO REQUIRED (EOI-1. ROPJ)

Cl WHILE EXECUTING STEPS Cl -22 1NROUGH C1-23:

!F THEN GAD TO COWTROL DV! CTR(Y. Al B ED r)MES LNAVAILABLE

APrx 8G RP PERl3llG CNNOT bE3rAJIUZED Q

MSRVs.P.E BEJI%GUSE 0 STA3IU RFV PRESPJDTHECONTNUOUS CONT1NUET STPC1-i2 PNEUMA.1IC SUFLY S LOSt LMLRGLNCI Rrv DLrRLDDJRI2AICN IS ROJRS)

AI1Y lNJCT O WRc: IEAUONLD C0NI1NtJFAT STrPC1Il WITHATLE%3T ft)MPFUNNING L

r.I-21 CAUTION 3

HPCI RCIC SUCT ON EMP A3OV 14d F L

• TLIZF V FEC NSRT4S lT NG VAJE MtHT.45 FCLLOW1D ETS.S.

OEPS3LJA1IO 3Y3TEM MSRA OlJ.yY.HSkSJPPR PL LVL lSA.E&5 FT Ill.

lPClWTH UCTIO rRDM ct r POlD..C lic HCICMTH UC11G1 FMCS l POlbLS I lb L

01-22 WICN V7ERLVLCROP3TO-2aOb.

2E!L CN11MJE L

Cl 23

UNIT 2 K TEA OOUMG IS OIJ1R EO-l. CP7)

WHILE EXECUTING STEPS C1-22 ThROUGH C1-23:

i ThEM OST CAD TO £PV CONTROL AIR EYiV COTL AIR VMIAaIJ kPFX 801 RPIJPICSS iM1Ir C80T si: STADIUZED Q

USRARE 8CI L6E0 TOSTABILJZE RPV PRESS AI(D TIL OJNTINL%IL$ GOt(TIIIUEATSTi:PC1 12 PILUMTIJPPLY ILVT i:MERGE lY R E.1PREiSUF5ZATKJ14 IS REOuIRi:D

,iY INJECTION SOURCL S AiJfM[O ONTINUIpTS1EPC1 wn rre ui.w nuiii L

C 1.21 CAUTION

1. 3 ELE1dATWSUPVR CIIABR PIS UkYTRIP IIC

1. 6 IPCIOR RCJC SUCTION TES# ABOVE 140 I L

Wr*SILIZS RPW PFSS NEARTHE EXISTBBG VALUE WrruFBE rA1 VTflA:

DgS4JRIzkT SYSTLL APPX OLYWIIE5 SUPPR PI LVUSABOVE 55 FT iSA tIPCI.WTII SUC11ON FROM CSTIFPOSSIBLC ISC RCICWflhI SLJCTIONFROL4 CsTIrposSIBLc isa L

C1-22 M:cRDRi95.

L r

UNIT 3 SRO Only Justification:

V STEAM COOLING IS REQUIRED (501-1, RCP-7 C-zo WHILE EXECUTING STEPS C1-22 THROUGH C1-23:

IF THEN

[Vs CCNTRCLAR BECCUES UNWAILABLE GROSSTIE CPJ 10 OW UUN I KOL P1K (APPX IF PRESS CAIIOTBS STSSIUZEO MYsPRE BEl1 1ED TO STASILIZE FF PRESS jTNE CON]1IIJOUS NUMATID SUPPLY IS LOST CONTINUE AT STEP CI 12

[>

EMEROENDY RPV DEPRESSURIZA11ON IS RECUIRSD PNY INJECTIDN SOLE ELIGN CONTINUE AT STCP Cl-IS

\MTH ATLEP& ONE PLPAP RU1M L

C1.21 CAU11ON

1. 3 EEVATB) SUPPRC-IMBF. PRESS MAY TRiP RCIC

1. 6 ICIolnOCnaTCMrPOOvCI4CT L

STABILIZE RR/ PRFSSNRARTHF FISTN(VI I MT-l THF POLLOwING SYSTCNS.

CFFSSII7ATlON SYSTFJ PPX MSRV0lLY WI-EN 3UPPR L LVL A5OE5.S T hA HCI WITh SUD11OI% FRDM CST IF POSS 815 11 C R;ICwrnSUDTIoN FROM CST IF POSSIBLE 118 L

Cl-fl I

C 1-23

J::1;::LvL0N L

SRO Only Justification: The SRO is required to have knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific subprocedures or emergency contingency procedures. The Minimum Zero Injection Reactor Water Level, MZIRWL value is a major transition point in the application of EOI Flowchart C-i, ALTERNATE LEVEL CONTROL. If this RPV water level is reached during steam cooling, the EOIs require transitioning to EOI-2, Emergency depressurization.

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, Le, how the system works, question flowpath, logic, component location?

Can the question be answered solely by knowing immediate operator actions? RO question Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?

Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

Does the question require one or more of the following?

• Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps SRO-only Knowledge of diagnostic steps and decision points in the estion EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures

• Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures No I Question might not be linked to 10 CFR 55.43(b)(5) for SRO-only

QUESTION 98 An individual who has already received an emergency TEDE dose of 23 rem has volunteered to accept another emergency exposure.

Which ONE of the following completes the statements?

He/she _( 1 )_ participate in this emergency exposure.

The _(2)_is (are) required to authorize an emergency exposure.

A. (1)CANNOT (2) Site Emergency Director ONLY B. (1)CANNOT (2) Site Emergency Director AND Radiation Protection Manager C. (1)CAN (2) Site Emergency Director ONLY D. (1)CAN (2) Site Emergency Director AND Radiation Protection Manager ANSWER: A

Level: RO SRO Tier# 3 Group#

Examination Outline Cross-Reference K/A#

G2.3.4 Importance Rating I

G2.3 .4 Knowledge of radiation exposure limits under normal or emergency conditions.

Explanation: A CORRECT: Part 1 is correct. This is considered a once-in-a-lifetime exposure, therefore he/she cannot participate. Part 2 is correct, Per EPIP-15 Appendix-B, the SED is the only required signature.

B- First Part: Correct. This is considered a once-in-a-lifetime exposure, therefore he/she cannot participate.

Second Part: Incorrect, Per EPIP- 15 Appendix-B, the SED is the y required signature.

C- First Part: Incorrect. This is considered a once-in-a-lifetime exposure, therefore he/she cannot participate.

Second Part: Correct.

D- First Part: Incorrect. This is considered a once-in-a-lifetime exposure, therefore he/she cannot participate since > 25 rem was received in the previous exposure. Second Part: Incorrect, Per EPIP-l 5 Appendix-B, the SED is the iy required signature.

Technical Reference(s): EPIP-1 5, Emergency Exposure Proposed references to be provided to applicants during examination: None Learning Objective (As available):

Question Source: Bank: X Modified Bank:

New:

Question History: Previous NRC: None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis:

10 CFR Part 55 Content: 55.43(b) 5 Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

3.3 Guidance for All Emergency Dose Limits 3.3.1 The exposure of personnel during emergency operations shall be maintained As Low As Reasonably Achievable (ALARA).

3.3.2 Internal exposure should be minimized by The use of respiratory protection equipment.

Respirator Protection Factors are provided in Standard Programs and Processes (SPP) 5.10. If a projected dose to the thyroid is expected to exceed 10 rem during emergency conditions, Potassium Iodide (Kl) should be issued. EPIP-14 contains information regarding issuance and precautions for the use of Kl.

3.3.3 Personnel undertaking an emergency operation in which the dose will exceed 10 CFR 20.1201 entitled Occupational Dose for Adults limits shall do so on a voluntary basis and with full awareness of the risks involved, including the numerical levels of dose at which acute effects of radiation will be incurred and numerical estimates of the risk of delayed effects as depicted on Appendix A. Acknowledgment of this decision shall be documented on Appendix B of this procedure by the individual involved in this activity.

3.3.4 Other factors being equal, older volunteers should be selected first.

3.3.5 Other factors being equal, selection of female volunteers capable of reproduction should be avoided.

3.3.6 Exposures under these conditions shall be limited to a once in a lifetime. Personnel who have received previous accident or emergency exposures in excess of 25 rem TEDE shall not participate in further emergency exposure assignments.

BROWNS FERRY EMERGENCY EXPOSURES EPI PI 5 APPENDIX B Page 1 of 1 ACKNOWLEDGMENT AND AUTHORIZATION TO EXCEED OCCUPATIONAL DOSE LIMITS READ THE FOLLOWING STATEMENT BEFORE SIGNING THIS FORM:

I acknowledge by signarnre on this fonu that I am volunteering for exposures in excess of 10 CFR 20J201 limits and that Ihave been made aware through training or a briefing of the risks involved, Briefing material was presented from Appendix A of this procedure.

The persons listed below have acknowledged and volunteered to receive dose limits in excess of 10CFR2O.1201 limits. Authorization is required by the Site Emergency Director to administer any emergency exposure limit. Authorization is acknowledged by Site Emergency Director signature on the bottom of this form.

Name Employee identification Signature Dose Limit (Please print Last. First, Ml) Number (EIN) (Rem)

Brief Description of Task:

Authorized by:

Site Emergency Director Time/Date LAST PAGE

SRO Only Justification: The SRO will need to know that EPIP-15, Emergency Exposures details the specifics regarding selection of a candidate for Emergency expose including the fact

( that the candidate must be a volunteer, and can only perform one emergency exposure per life time. Additionally the SRO would need to know that the form authorizing exceeding occupational dose limits is approved by the Site Emergency Director.

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. Le, how the system works, Yes RO ques ion flowpath. logic, component location?

Can the question be answered solely by knowing I immediate operator actions? Yes RO question Can the question be answered solely by knowing entry conditions for AOPs or plant parameters Yes RO question that require direct entry to major EOPs?

Can the question be answered solely by knowing the purpose, overall sequence of events, or Yes RO question overall mitigative strategy of a procedure?

Does the question require one or more of the following?

• Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed

• Knowledge of when to implement attachments and appendices including how to coordinate these items with procedure steps 0 SROonly question

• 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 No

[Question might not be linked to CFR 55A3(b)(5) for SRO-only

QUESTION 99 Unit I is currently in Mode 1. Fuel movement is in progress in Unit 1 Spent Fuel Pool.

During the movement of irradiated fuel, a fuel bundle is severely damaged, resulting in the following conditions.

• Field Assessment Team surveys at the Site Boundary Indicate Iodine-131 levels have exceeded the General Emergency limit.

• Radiation levels in numerous areas of the Reactor Building are above Max Safe.

Which ONE of the following completes the statements?

In accordance with the EOIs the Unit Supervisor must _(1)_.

In accordance with the EOI Program Manual the basis for this action is to _(2)_.

Note: 1 -EOI- 1, RPV Control 1 -EOI-4, Radioactive Release Control 1-GOI-100-12A, Unit Shutdown from Power Operation to Cold Shutdown and Reductions in Power During Power Operations A. (1) enter 1-GOI-100-12A, from 1-EOI-3 and proceed to Cold Shutdown (2) mitigate a direct and immediate threat, relative to plant equipment and personnel, both on and off site B. (1) enter 1-GOI-100-12A, from 1-EOI-3 and proceed to Cold Shutdown (2) limit the release of radioactivity discharging into areas outside the primary and secondary containments C. (1) enter 1 -EOI- 1 from 1 -EOI-4, Scram the Reactor and Emergency Depressurize (2) mitigate a direct and immediate threat, relative to plant equipment and personnel, both on and off site D. (1) enter 1 -EOI-1 from 1 -EOI-4, Scram the Reactor and Emergency Depressurize (2) limit the release of radioactivity discharging into areas outside the primary and secondary containments ANSWER: A

Level: RO SRO Tier# 3 Group#

Examination Outline Cross-Reference G2.4.18 Importance Rating f 4.0 G2.4. 18 Knowledge of the specific bases for EOPs.

Explanation: Answer A CORRECT: The US must determine that Emergency Depressurization will not

reduce radiation levels in secondary containment and with two radiation levels in the Reactor building above max safe, Cold Shutdown is required. The basis for this action is to mitigate a direct and immediate threat, relative to plant equipment and personnel, both on and off site.

B Incorrect First part is correct. Second part is plausible in that this is the correct basis for Emergency Depressurization.

C Incorrect First part incorrect. The candidate may determine this is the correct action based on meeting the entry requirements for EOI-4 and meeting the Reactor Scram and Emergency Depressurization requirement if a primary system was discharging. Second part is the correct basis for proceeding to cold shutdown.

D Incorrect First part incorrect. The candidate may determine this is the correct action based on meeting the entry requirements for EOI-4 and meeting the Reactor Scram and Emergency Depressurization requirement if a primary system was discharging. Second part is plausible in that this is the correct basis for Emergency Depressurization.

Technical Reference(s): l-EOI-3, Secondary Containment Control 0-EOI-4, Radioactive Release Control, EOI Program Manual Proposed references to be provided to applicants during examination: None Learning Objective (As available):

Question Source: Bank: X Modified Bank:

New Question History: Previous NRC: No Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Bases for 1-EOI-3 steps SCIR-9 and 10 This decision step has the operator evaluate status of radiation levels in all secondary containment areas listed in Table 4, to determine if a normal reactor shutdown is required.

To reach this step, it was previously determined that no primary systems were discharging into secondary containment. When this condition exists, and when radiation levels in two or more secondary containment areas exceed their respective maximum safe operating values, it is prudent to commence aii orderly reactor shutdown because a direct and immediate threat exists relative to plant equipment and to personnel, both on and off site. The operator continues in this procedure at Step SC/R-1 0 where actions to perform a normal plant shutdown are directed.

L J

L L

SRO Only Justification: The SRO will assess plant conditions and recognize that no primary systems were discharging into secondary containment and emergency depressurization will not reduce discharge into the secondary containment. Additionally, the SRO will need to know that EOI-3 Secondary Containment Control is the procedure with which to proceed and it will direct a Normal Shutdown per GOI-lOO-12A based on these conditions. The SRO will also need to know the EOI Program Manual the basis for this action.

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, Le, how the system works, flowpath. logic, component location?

ltion Can the question be answered solely by knowing immediate operator actions? Yes I RO question I1 I Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?

Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

I Does the question require one or more of the following?

Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

• 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 No Question might not be linked to I 10 CFR 5543(b)(5) for SRO-only

QUESTION 100 Which ONE of the following completes the statement regarding the proper EOI Flow Chart direction for entry into a Contingency Procedure and the impact of the entry on the use of the original (directing) procedure?

Entry into a Contingency Procedure by the EOI Flowcharts will (1) and, upon entry to the Contingency Procedure, the Unit Supervisor (2) exit the original (sending) procedure.

A. (1) be directed in an override OR an exit arrow to the Contingency (2) will ALWAYS B. (1) be directed in an override OR an exit arrow to the Contingency (2) may or may not C. (1) be directed in an override with an exit arrow to the Contingency in the override (2) will ALWAYS D. (1) be directed in an override with an exit arrow to the Contingency in the override (2) may or may not Answer: B

Level: RO SRO Tier# 3 Group#

Examination Outline Cross-Reference KIA#

G2.4. 19 Importance Rating 4.1 Knowledge of EOP layout, symbols, and icons.

Explanation: B CORRECT: First Part: For example in EOI-l, entry into C-4 is directed as an override RCIL-3 and entry into C-i in RCIL-13 with a match arrow. Second Part: Entry into C-2 from C-i directs the operator to exit RCIP but not to exit C-i.

A Incorrect Plausible as part (1) is correct but as noted above the originating procedure may or may not be exited.

C Incorrect Plausible Match lines are used but not in conjunction with the overrides.

D Incorrect Plausible Match lines are used but not in conjunction with the overrides.

Technical Reference(s): EOIPM 0-VIII-A Proposed references to be provided to applicants during examination: None Learning Objective (As available):

Question Source: Bank:

Modified Bank:

New: X Question History: Previous NRC: None Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.43 b (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

EOIPM O-VIII-A When the operator reaches an Exit Arrow the flowpath is exited at that point and operation continues in the procedure designated by the Exit Arrow. When the target procedure is entered the operator locates the Entry Arrow displaying the procedure number and step number, or matchmark letter, associated with the Exit Arrow.

WHILE EXECUTING THE FOLLOWING STEPS:

IF THEN It has NOT been determined that EXIT RC/L and the reactor will remain subcritical ENTER CS, Level/Power Control without boron under all conditions RPV water lvi CANNOT be EXIT RCIL and determined ENTER 04, RPV Flooding PC water lvi CANNOT be maintained below 105 ft STOP inj into the RPV from sources OR external to the PC NOT required for adequate core cooling Suppr chmbr press CANNOT be maintained below 55 psig L

RC/L-3 WHILE EXECUTING THIS PROCEDURE:

IF THEN

. EXIT RCIP and Steam CooIirg Emergency RPV depresurization ENTER C2. Emergency RPV required Depressuthacn It has NOT been deteimined that a

EXIT this procedure and the reactor wII remain subontical ENTER c5. LevellPowerContro4 without boron under all condbons RPV water M CANNOT e xrr this procedure and determined ENTER C4, RPV Flooding PC water lvi CANNOT be rnairained below 1 ft STOP inj into the RPV from sources OR external to the PC NOT requiTed for adequate core cooling Suppr chmbr press CANNOT maintained below 5 psig L

cl-I

1 Oi-i. RPV Control at Step RciPL _/

\

L Can NOL RPV water lvi be restored and maAntained above -162 1n cl-i L

250 j 1

230[

SRO Only Justification: The SRO will need to know the proper usage of flow chart marking in the EOIs. This is the responsibility of the SRO to drive the flowcharts and direct operators according the flowchart usage rules.

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, question flowpath, logic, component location?

Can the question be answered solely by knowing immediate operator actions?

Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs?

Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure?

Ii Does the question require one or more of the following?

Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed

• Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps oy

• 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 No Question might not be linked to 10 CFR 5543(b)(5) for SRO-only