Text

Feb-March 05000259/2008301 Exam Draft SRO Written Exam (Part 1 of 2)
	ML081370271
Person / Time
Site:	Browns Ferry
Issue date:	04/08/2008
From:	; NRC/RGN-II
To:	; Tennessee Valley Authority
References
	50-259/08-301
	Download: ML081370271 (106)
	v • d • e

{{#Wiki_filter:Draft Submittal (Pink Paper) Senior Reactor Operator Written Exam -:[j;en w;.)s kJ!~Y olt)tJ6---~O/

ANSWER KEY REPORT for 0610 NRC SRO Exam Test Form: 0 SRO 295006AA2.01 1 D SRO 295021AA2.07 1 A 3 SRO 295024G2.1.28 1 D 4 SRO 295026EA2.01 1 D 5 SRO 295037EA2.05 1 C 6 SRO 295038G2.2.22 1 A 7 SRO 600000G2.2.24 1 C 8 SRO 29500202.4.6 1 B 9 SRO 295009G2.4.31 1 C 10 SRO 500000EA2.02 1 D 11 SRO 203000A2.16 1 C 12 SRO 215003G2. 1.14 1 C 13 SRO 259002G2.1.23 1 C 14 SR020S000A2.06 1 A 15 SRO 212000A2.12 1 D 16 SRO 268000A2.01 1 D 17 SRO 271000G2.4.36 1 C 18 SRO 288000A2.03 1 B 19 SRO GENERIC 2.1.12 1 C 20 SRO GENERIC 2.2.22 1 D 21 SRO GENERIC 2.2.24 1 A SRO GENERIC 2.3.3 1 D SRO GENERIC 2.3.9 1 A SRO GENERIC 2.4.21 1 D 25 SRO GENERIC 2.4.30 1 B Saturday, December 22, 2007 3:38:26 PM 1

( ( ( 1. SRO 29S006AA2.01 OOllC/AfflGlIRPSI129S006AA2.0l/4.6/SRO ONLY/11l27/07 RMS Given the following plant conditions:

Unit-1 was at 100% RTP with RPS "A" on Alternate due to problems with the MG set.

A fault occurred on 1B 480V RMOV Board causing a complete loss of RPS and reactor scram.

The following conditions are observed: - Multiple control rods failed to insert on the scram - PRNM indication is unavailable - Reactor pressure is being maintained at 900 psig and steady with 3 MSRVs open - EOI-1 and C-5 have been entered Which ONE of the following describes the approximate value of reactor power and the appropriate operator response? REFERENCE PROVIDED A. Power is approximately 12%. Maintain RPV level +2 to +51 inches untillRMs are inserted and on-scale. B. Power is approximately 20%. Maintain RPV level +2 to +51 inches until IRMs are inserted and on-scale. C. Power is approximately 12%. Stop and prevent injection in accordance with EOI Appendix 4. D!' Power is approximately 20%. Stop and prevent injection in accordance with EOI Appendix 4. KIA Statement: 295006 SCRAM 11 AA2.01 - Ability to determine and/or interpret the following as they apply to SCRAM: Reactor power KIA Justification: This question satisfies the KiA statement by requiring the candidate to use specific plant conditions to determine the reactor power level using alternate methods and select the appropriate procedure to address those conditions. Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CfR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. 0610 NRC SRO Exam

( ( REFERENCE PROVIDED: 1-EOI-C5 FLowchart Plausibility Analysis: In order to answer this question correctly, the candidate must: Detennine the relationship between MSRV steam flow and reactor power as follows: Total rated steam flow through 13 MSRVs = 84% divided by, 13 MSRVs=6 .5% multiplied by 3 open MSRVs = 19.5% total. This becomes confusing for a n vice operator candidate due to the difference between steam flow through an MSRV and steam flow through an open Bypass Valve (BPV). Detennine that it is not appropriate to wait untillRMs are inserted to obtain an instrumentation reading for power level detennination prior to perfonning level control actions, but it IS required to wait until IRMs are inserted to determine if the reactor is sub-critical under all conditions without boron injection. This is an occasional point of confusion among novice operator candidates. A is incorrect based on power assiciated with 3 BPVs open and not 3 MSRVs. Also incorrect based on waiting to perform the required actions using reason #2 above. B is incorrect based on waiting to perform the required actions using reason #2 above. C is incorrect based power assiciated with 3 BPVs open and not 3 MSRVs. #1. D is Correct

( (6) The worst over pressure transient is: (a) 3-second closure of all MSIVs neglecting the direct scram (valve position scram). (b) Results in a maximum vessel pressure which, if a neutron flux scram is assumed considering 12 valves operable, results in adequate margin to the code allowable over pressure limit of 1375 psig bottom head pressure. (7) To meet operational design, the analysis of the plant isolation transient (generator load reject without bypass valves) shows that 12 of the 13 valves limit peak pressure to a value well below the limit of 1375 psig. b. The total safety I relief valve capacity has been established to meet the over pressure protection criteria of the ASME code. (1) There are 13 Safety I Relief valves. (a) Each SRV has a capacity of 905,OOOlblhr @ 1135psig. This gives a total capacity ~ 84.1% (79.5% EPU) design steam flow at the reference pressure. (b) Valve leakage is detected by a temperature element and an acoustic monitor on each tailpipe. However, only the acoustic monitor will generate an alarm on panel 9-3. OPL171.009 Revision 10 Page 15 of 62 Obj. V.B.6 Obj. V.C.4

/'\\ ( ( 2. SRO 295021AA2.07 OOIIMEMffIlGlIRHRISDCIB31/295021AA2.071ISRO ONLY/ll129/07 RMS Unit-2 is in a refueling outage and is partially defueled when a loss of shutdown cooling occurs and cannot be re-established. Which ONE of the following cautions must be satisfied to operate a reactor recirculation pump to establish forced cooling flow in compliance with 2-AOI-74-1, Loss of Shutdown Cooling? A. v Ensure the in-core instrumentation tubes are protected on all sides by control rod blade guides or fuel bundles. B. Ensure core flow rate is less than 4000 gpm to prevent ejecting control rod blade guides. C. Maintain recirculation pump at minimum speed to to reduce hydraulic forces and vibration stresses on jet pumps and retainers. D. Ensure core flow rate is less than 4000 gpm to prevent the addition of excessive heat to the primary coolant. KIA Statement: 295021 Loss of Shutdown Cooling I 4 AA2.07 - Ability to determine and/or interpret the following as they apply to LOSS OF SHUTDOWN COOLING: Reactor recirculation flow KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the actions necessary to establish recirculation flow following a Loss of Shutdown Cooling with specific emphasis placed on refueling activities. References: 2-AOI-74-1 and 2-01-68 Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of information. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (7) Fuel handling facilities and procedures. 0610 NRC SRO Exam

( REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly, the candidate must: Demonstrate knowledge of precautions, limitations and embedded Cautions and Notes in the Reactor Recirculation System procedure (2-01-68) that directly apply to starting a Recirc Pump during refueling activities. The distractors provided are based on existing limitations with incorrect applicability statements in the distractors. The associated Iimtations from 2-01-68 are attached for comparison. A is the correct answer. B is incorrect based on the required flow. The actual flow limitation is 34,000 gpm. C is incorrect based on the cause of hydraulic forces. The correct basis is unbalanced Jet Pump flows, not recirc speed. D is incorrect based on the applicability of this NOTE. Adding excessive heat to the coolant is only justification for removing ONE recirc pump from service, not establishing forced coolant flow when required by procedure.

BFN Unit 2 Reactor Recirculation System ( 3.0 PRECAUTIONS AND LIMITATIONS (continued) J. The following recirc pump motor limitations should be observed: 1. Starts should be limited to two starts per hour from ambient temperature and one start per hour from rated temperature. 2. Recirc pump motor SHALL NOT be operated when RBCCW cooling water inlet temperature exceeds 105°F. 3. The recirc pump should NOT be operated in parallel with RHR pumps on the same recirc loop. 4. During refueling operations, the recirc pumps must remain shutdown to ensure in-core instrumentation detector vibration protection, unless the in-core instrumentation detectors are protected on all sides by control rod blade guides or fuel bundles. 5. After operation of 4160V breakers, the charging springs should be verified recharged. K. A recirc pump may NOT be started unless the temperature of the coolant between the dome and the bottom head drain are within 145°F of each other. If this limit is exceed and a cooldown is NOT started and continued, the pressure limit for the reactor vessel bottom head will be exceeded. Recirc pump restart when the reactor is pressurized is allowed ONLY if temperature indication is available. L. [NER/C] When the reactor is operating with a single recirc pump in service, observe the following: 1. When a recirc pump is shut down (or shuts down), the discharge valve should be closed for 5 minutes to ensure that the pump comes to a complete stop and prevent reverse rotation. 2. Idle recirc pump discharge valve should be opened after the initial 5 minute closure to maintain idle recirc loop temperatures within associated temperature limits. 3. As long as recirc loop suction or discharge is open, maintain recirc pump seal purge to prevent thermal stress on the seal cavity. 4. The operating recirc pump flow should be maintained less than 46,600 gpm. 5. Core flow indication may be considered accurate above 41 x 1061bmlhr (2-FI-68-46 and 2-FI-68-48).

BFN Reactor Recirculation System 2-01-68 Unit 2 Rev. 0124 Page 16 of 166 ( 3.0 PRECAUTIONS AND LlMITAnONS (continued) U. [lIfC) ICS Core Thermal Heat Balance inputs on Illustration 6 should be monitored periodically, and at any time reactor power changes are performed. This ensures these inputs are in agreement and prevents exceeding the license limit. The following requirements must be met: 1. Feedwater temperature points on Illustration 6 should NOT differ by more than 2 degrees. A difference between any feedwater temperature points of greater than 2 degrees requires the notification of System Engineering and suspending the raising of power until the discrepancy is resolved. 2. HI or HI-HI alarm setpoints on Illustration 6 should NOT be exceeded. An alarm setpoint being exceeded requires notifying the Shift Manager immediately and, if action cannot be taken immediately to return the value to within limits, System Engineering is to be notified for assistance [BFPER951914). V. Recirc Flow changes made during the later part of the operating cycle (Coastdown) could cause core flow values to approach or exceed the allowable values of the Increased Core Flow (ICF) Region of the power to flow map. Instruments used to monitor pump speed and core flow should be identified in the Reactivity Control Plan. These values should be recorded prior to reducing core flow and used as a benchmark to reestablish the previous conditions when returning to power. Increased caution should be used when changes in Recirc Flow are made in this area. W. During refueling operation, to prevent blade guide ejection whenever any fuel bundle is removed, core flow is required to be limited to less than 34,000 gpm, or core D/P to less than 4 psid. X. TS BASES SR 3.5.1.5, If Recirc Pump 2A(2B) Discharge Valve, 2-FCV-068-0003(0079), is declared inoperable while the valve is open, the associated LPCI Subsystem must be declared INOPERABLE. Y. TS SR 3.4.2.1, 2-SR-3.4.2.1, Jet Pump Mismatch and Operability, is required to be performed 24 hours after reaching> 25% RTP and/or within 4 hours after returning an idle recirc pump to service. Z. 2-SR-3.4.9.3&4, requires the recirc pump be started within 15 minutes of the performance of this SR. AA. During the warm-up of an idle recirc loop prior to starting the idle recirc pump, the loop heatup rate is limited to 90°Flhr. BB. [lIfC) When initiating manual runbacks, the appropriate manual pushbutton must be depressed until the backlight is blinking, then the pushbutton can be released.[PER 98-013557-000)

( BFN Reactor Recirculation System 2-01-68 Unit 2 Rev. 0124 Page 55 of 166 7.0 SHUTDOWN 7.1 Recirc Pump Shutdown (Plant NOT in Mode 1) CAUTIONS 1) Recirc System operation is restricted by criteria in Unit 2 Power to Flow Map (ICS or Station Reactor Engineering, 0-TI-248) and Illustration 1. 2) The Recirc System should be operated such that recirc loops with forced flow have balanced jet pump flows to reduce hydraulic forces and vibration stresses on jet pumps and retainers. NOTES 1) Section 7.2 provides instructions for stopping a recirc pump during power operation. 2) All operations are performed from Panel 2-9-4 unless noted otherwise. 3) One pump may be removed from service to conserve power, minimize heat input into the Reactor, for maintenance or for testing. 4) The Unit Supervisor may authorize the removal of one recirc pump from service while cooling down. 5) When depressing the switches which control the Recire Drives, these switches must be firmly depressed to ensure all the contacts are made-up. [1] REVIEW Section 3.0, Precautions and Limitations. D [2] VERIFY in service recirc pump(s) are operating at 345 RPM to 480 RPM pump speed. D [3] MAINTAIN at least one recire pump operating at minimum speed (approximately 345 RPM to 480 RPM) until the shutdown cooling system is flushed and ready for operation or until the Reactor is at the desired shutdown condition. D

( 3. SRO 295024G2.1.28 00 llMEMlflGlIBASIS//295024G2.1.28//SRO ONLY111/29/07 RMS EOI-2, Primary Containment Control, section PC/P directs the operator to vent the primary containment irrespective of radioactivity releases before containment pressure rises to 55 psig. Which ONE of the following describes the basis for the maximum primary containment pressure limit at or below 55 psig in the EOI's? A. It is the maximum pressure capability of the containment structure. B. It is the maximum pressure that MSRVs can be opened and remain open. C. The RPV vent valves may not be capable of being opened and closed above this pressure to vent the RPV for containment flooding. D.'" The primary containment vent valves may not be capable of being opened and closed above this pressure to reject decay heat. KIA Statement: 295024 High Drywell Pressure I 5 2.1.28 - Conduct of Operations Knowledge of the purpose and function of major system components and controls. KIA Justification: This questlon .satlstles the KIA statement by requiring the candidate to demonstrate knowledge of the basis for primary containment high pressure limits. References: FSAR limitations described in EOIPM Section O-V-D page 55 of 244 Level of Knowledae Justification: This question is rated as MEM due to the requirement to recan or recognize discrete bits of information. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (1) Conditions and limitations in the facility license. 0610 NRC SRO Exam

( REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly, the candidate must: Demonstrate knowledge of the bases considered for determination of the Primary Containment pressure limit and the specific limitation applicable to BFN. This limitation is provided in greater detail in the BFN Final Safety Analysis Report (FSAR), however the bases associated with the distractors are outlined in the EOI Program Manual and not in the FSAR. The limitation is defined as the lesser of four possible limitations. The distractors are each of the three limitations not applicable to BFN. Plausibility for the distractors is assured based on their potential applicability. A is incorrect. Item #1 B is incorrect. Item #3 C is incorrect. Item #4 o is correct. Item #2 EOIPM Section O-V-D page 55 of 244 is attached for reference.

EOI-2, PRIMARY CONTAINMENT CONTROL BASES EOI PROGRAM MANUAL SECTION O-Y-D DISCUSSION: STEP PCIP-13 (~ 1 1 This actionstep directs the operatoris to useall availablecontainmentpressurecontrolsystemsto maintainsuppression chamberpressurebelowPrimary ContainmentPressureLimit Engineering calculations havedetermined that primarycontainment integritycanbeassured,and coredamage(that mightbecausedby inabilityto vent theRPVto permit injectioninto the RPV) ispreventedwhensuppressionchamberpressureismaintainedbelow<A.61>, Primary Containment PressureLimit This valueis roundedoffin the EOIto use the closest,most conservative valuethatcanbeaccuratelydetermined onavailableinstrumentation. Primary ContainmentPressureLimitis definedto be the lesserofeither: 1)pressurecapability of the containment, or 2) maximumcontainment pressureat which ventvalvescan be openedand closedto rejectall decayheat fromprimarycontainment, or 3) maximumprimarycontainment pressureat whichMSRVs canbeopenedand willremain open,or4) maximumcontainment pressureat whichventvalvescanbe openedand closedto ventthe RPV.At BFN,this limitis a functionofprimary containmentventvalveoperability(item.2). Themaximumcontainment pressurethat primarycontainmentventvalvescan be openedand closedis <A.61>. The operatoris then directed to followthe branch designator,~,to Step PCIP-14.

'. REVISION 0 PAGE 55 OF 244 SEcnON o-V-D

( 4. SRO 295026EA2.01 OOIlC/AfI'IIGlIE0I-211295026EA2.01/lSRO ONLY/l1l29107 RMS Given the following plant conditions:

Unit 2 is in Mode 3 preparing for a plant startup per 2-GOI-1 0Q-1 A, "Unit Startup and Power Operation".

Surveillance 2-SR-3.5.1.8, "HPCI Main and Booster Pump Set Developed Head and Flow Rate Test at 150 psig" is in progress using auxiliary steam.

The following conditions exist due to sustained operation of HPCI:

SUPPR POOL AVERAGE TEMP HIGH (3E-W12) in alarm

Both loops of suppression pool cooling are in service

Suppression Chamber Water Temperature Check, 2-SR-3.6.2.1.1 in progress with Panel 2-9-3, SUPPR POOL BULK TEMP, 2-TI-64-161, and SUPPR POOL BULK TEMP, 2-TI-64-162 reading 95°F. Which ONE of the following describes the required actions in accordance with Emergency Operating Instructions and Technical Specifications? REFERENCE PROVIDED A. Do NOT enter EOI-2 Verify < 1050F and log every 5 minutes until the heat addition is terminated. B. Enter EOI-2 Suspend all testing that adds heat to the suppression pool. c. Do NOT enter EOI-2 Verify 110°F once per hour D." Enter EOI-2 No Tech Spec actions required.

( ( KIA Statement: 295026 Suppression Pool High Water Temp. 15 EA2.01 - Ability to determine and/or interpret the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Suppression pool water temperature KIA Justification: This question satisfies the KIA statement by requiring the candidate to interpret plant indications of Suppression Pool temperature and determine the appropriate procedures required to address those conditions. References: 2-EOI-2, U2 Tech Spec Section 3.6.2.1 Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. 0610 NRC SRO Exam REFERENCE PROVIDED: U2 Tech Spec Section 3.6.2.1 Plausibility Analysis: In order to answer this question correctly, the candidate must: Determine that an EOI entry condition has been reached and EOI entry is required regardless of the MODE of operation. Not entering EOls becomes plausible if the candidate correctly determines that the 95°F Tech Spec limit does not apply under these conditions or determines that EOI actions are already being performed by having Suppression Pool Cooling in service, so EOI entry will provide little value. Determine the applicability of the 95°F Tech Spec limit based on given plant conditions. The distractors are based on other actions associated with adding heat to the Suppression Pool. A is incorrect based on the ARP requirements for SUPPR POOL AVERAGE TEMP HIGH (3E-W12) which states, IF alarm is valid, THEN ENTER 2-EOI-2 Flowchart. B is incorrect based on the temperature (>105°F) which mandates that action, has not yet been exceeded. THe EOI entry is correct. C is incorrect based on the ARP requirements for SUPPR POOL AVERAGE TEMP HIGH (3E-W12) which states, IF alarm is valid, THEN ENTER 2-EOI-2 Flowchart. The Tech Spec action is appropriate. D is correct.

-- --------------------------- 2-ARP-9-3E Unit 2 2-XA-55-3E Rev. 0021 __________________________________________________1'8Q!t__t5of ~ _ (Alarm re-f1ashes after each rise in temperature) SUPPR POOL AVERAGE TEMP HIGH 2-TA-64-162A [12 (Page 1 of 1) SensorfTrip Point: DIVI TS-64-161J TS-64-161K DIV II TS-64-162J TS-64-162K 95°E Hi 110°F Hi-Hi 120°F Hi-Hi-Hi Sensor Location: Probable Cause: Automatic Action: Panel 9-87 and 9-88 Aux. Inst. Room, EI 593'. A. HPCI and/or RCIC in service or testing. B. Leaking or stuck open SRV, or testing of SRVs. C. Temperature element failure. D. Loss of both Div. I ECCS ATU inverters or both Div. II ECCS ATU inverters. None Operator Action: A. IF alarm is valid, THEN ENTER 2-EOI-2 F.lowchart. B. VERIFY HPCI and RCIC are in standby readiness AND Main Steam SRVs are NOT lifting. C. VERIFY if a TE has failed or is reading high by selecting each element on SUPPR POOL SELECTED TEMP recorder/indicator, 2-TRfTI-64-161 and-162. D. NOTIFY Shift Manager. E. IF recorderlindicator has lost power, THEN DISPATCH personnel to: 1. CHECK 250V DC power supply to ECCS ATU inverters. a. Div. I 250V DC Rx MOV Bd. 2B, Breaker 8A. (RxBldg, E1593', R-14 Q-L1NE) b. Div. II 250V DC Rx MOV Bd. 2A, Breaker 11A1. (RX Bldg, EI621', R-14 R-L1NE) 2. RESET and CLOSE ECCS ATU inverter DC input breaker and then CLOSE the AC output breaker. a. Div. I ECCS ATU inverter, EI593', Shutdown Board Room D, R-14 Q-L1NE. b. Div. II ECCS ATU inverter EI621', Shutdown Board Room C, R-14 P-L1NE. o o oo oo o o o o o References: F. REFER TO Tech Spec 3.6.2.1. 2-45E620-1 2-47E610-64-3 Technical Specifications 3.6.2.1 and 5.4, 5.5 2-45E712-1,2 FSAR Section 13.0 o

Suppression Pool Average Temperature 3.6.2.1 3.6 CONTAINMENT SYSTEMS 3.6.2.1 Suppression Pool Average Temperature LCO 3.6.2.1 Suppression pool average temperature shall be: a. s 950F when any OPERABLE intermediate range monitor (IRM) channel is > 70/125 divisions of full scale on Range 7 and no testing that adds heat to the suppression pool is being performed; b.

105°F when any OPERABLE IRM channel is> 70/125

divisions of full scale on Range 7 and testing that adds heat to the suppression pool is being performed; and c.

110°F when all OPERABLE IRM channels are s 701125

divisions of full scale on Range 7. ( APPLICABILITY: MODES 1, 2, and 3. BFN-UNIT2 3.6-24 Amendment No. 253

ACTIONS Suppression Pool Average Temperature 3.6.2.1 CONDITION REQUIRED ACTION COMPLETION TIME A. Suppression pool A.1 Verify suppression pool Once per hour average temperature average temperature > 95°F but s 110°F. s 110°F. AND AND Any OPERABLE IRM A.2 Restore suppression pool 24 hours channel> 70/125 average temperature to divisions of full scale on ~ 95°F. Range 7. AND Not performing testing that adds heat to the suppression pool. B. Required Action and B.1 Reduce THERMAL 12 hours associated Completion POWER until all Time of Condition A not OPERABLE IRM met. channels are s 70/125 divisions of full scale on Range 7. (continued) l ..-- .. BFN-UNIT2 3.6-25 Amendment No. 253

Suppression Pool Average Temperature 3.6.2.1 CTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. Suppression pool C.1 Suspend all testing that Immediately average temperature adds heat to the > 105°F. suppression pool. AND Any OPERABLE IRM channel> 70/125 divisions of full scale on Range 7. AND Performing testing that adds heat to the suppression pool. D. Suppression pool D.1 Place the reactor mode Immediately average temperature switch in the shutdown > 110°F but s 120°F. position. AND D.2 Verify suppression pool Once per 30 average temperature minutes ~ 120°F. AND D.3 Be in MODE 4. 36 hours A ( (continued) BFN-UNIT 2 3.6-26 Amendment No. 253

-: \\ ACTIONS (continued) Suppression Pool Average Temperature 3.6.2.1 CONDITION REQUIRED ACTION COMPLETION TIME E. Suppression pool E.1 Depressurize the reactor 12 hours average temperature vessel to < 200 psig. > 120°F. AND E.2 Be in MODE 4. 36 hours ( BFN-UNIT2 3.6-27 Amendment No. 253

/ SURVEILLANCE REQUIREMENTS SURVEILLANCE Suppression Pool Average Temperature 3.6.2.1 FREQUENCY SR 3.6.2.1.1 Verify suppression pool average temperature 24 hours is within the applicable limits. 5 minutes when performing testing that adds heat to the suppression pool ( BFN-UNIT2 3.6-28 Amendment No. 253

E MINATION REFERENCE PROVIDED TO .CANDIDATE

Suppression Pool Average Temperature 3.6.2.1 3.6 CONTAINMENT SYSTEMS 3.6.2.1 Suppression Pool Average Temperature LCO 3.6.2.1 Suppression pool average temperature shall be: a. s 95°F when any OPERABLE intermediate range monitor (IRM) channel is > 70/125 divisions of full scale on Range 7 and no testing that adds heat to the suppression pool is being performed; b. s 105°F when any OPERABLE IRM channel is> 70/125 divisions of full scale on Range 7 and testing that adds heat to the suppression pool is being performed; and c. s 110°F when all OPERABLE IRM channels are s 70/125 dh(isions of full scale on Range 7. ( APPLICABILITY: MODES 1, 2, and 3. BFN-UNIT 2 3.6-24 Amendment No. 253

/ ACTIONS Suppression Pool Average Temperature 3.6.2.1 CONDITION REQUIRED ACTION COMPLETION TIME A. Suppression pool A.1 Verify suppression pool Once per hour average temperature average temperature > 95°F but s 110°F. s 110°F. AND AND Any OPERABLE IRM A.2 Restore suppression pool 24 hours channel> 70/125 average temperature to divisions of full scale on s 95°F. Range 7. AND Not performing testing that adds heat to the suppression pool. B. Required Action and B.1 Reduce THERMAL 12 hours associated Completion POWER until all Time of Condition A not OPERABLE IRM met. channels are s 70/125 divisions of full scale on Range 7. (continued) BFN-UNIT2 3.6-25 Amendment No. 253

Suppression Pool Average Temperature 3.6.2.1 CTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. Suppression pool C.1 Suspend all testing that Immediately average temperature adds heat to the > 105°F. suppression pool. AND Any OPERABLE IRM channel> 70/125 divisions of full scale on Range 7. AND Performing testing that adds heat to the suppression pool. D. Suppression pool 0.1 Place the reactor mode Immediately average temperature switch in the shutdown > 110°F but s 120°F. position. AND 0.2 Verify suppression pool Once per 30 average temperature minutes s 120°F. AND 0 .3 Bein MODE 4. 36 hours A ( (continued) BFN-UNIT2 3.6-26 Amendment No. 253

Suppression Pool Average Temperature 3.6.2.1 ( ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME E. Suppression pool E.1 Depressurize the reactor 12 hours average temperature vessel to < 200 psig. > 120°F. AND E.2 Bein MODE 4. 36 hours BFN-UNIT2 3.6-27 Amendment No. 253

( SURVEILLANCE REQUIREMENTS SURVEILLANCE Suppression Pool Average Temperature 3.6.2.1 FREQUENCY ( SR 3.6.2.1.1 Verify suppression pool average temperature 24 hours is within the applicable limits. 5 minutes when performing testing that adds heat to the suppression pool BFN-UNIT2 3.6-28 Amendment No. 253

( 5. SRO 295037EA2.05 OOllC/AffllGl/EOI-II/295037EA2.051ISROONLY/l1l29/07 RMS Given the following plant conditions:

Unit-1 reactor has scrammed automatically due to a load reject resulting from an improper switching order executed several miles from the plant.

The following indications are present on control room Panel 9-5:. - The left side of the full core display shows green control rod positions reading Notch 00. - The right side of the full core display shows blank control rod position indications. Which ONE of the following describes the probable cause of the full core display indications and which EOI appendicies are required to restore all control rod position indications? A. RPS Scram Group A2. and A3 failed to de-energize. Perform EOI AppendiX 1A, "Removal and Replacement of RPS Scram Solenoid Fuses" or EOI Appendix 1B, "Venting and Repressurizing the Scram Pilot Air Header" . B. RPS Scram Group A1 and A4 failed to de-energize. Perform EOI Appendix 2, "Defeating ARI Logic Trips" and EOI Appendix 1F, "Manual Scram". C~ West SDV is hydraulicly locked. Perform EOI Appendix 2, "Defeating ARI Logic Trips" and EOI Appendix 1F, "Manual Scram". D. East SDV is hydraulicly locked. Perform EOI Appendix 1A, "Removal and Replacement of RPS . Scram Solenoid Fuses" or EOI Appendix 1B, Venting and Repressurizing the Scram Pilot Air Header". KIA Statement: 295037 SCRAM Condition Present and Power Above APRM Downscale or Unknown 11 EA2..05 - Ability to determine and/or interpret the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Control rod position KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the appropriate actions to restore control rod position indication during an ATWS. References: 1-EOI-1, 1-EOI Appendicies 1F, 2 Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. 0610 NRC SRO Exam

( REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly, the candidate must: Demonstrate a detailed knowledge of the operation of the CRD hydraulic system under off-nonnal conditions including the following : 1. The relationship between RPS Scram Groups (A1 - A4) and the control rods which are assigned to each specifc group. 2. Knowledge that any RPS Scram Group which inserts will initiate a full-scram due to SDV High Level. 3. The relationship between the EOI Appendicies and their effect on the CRD Hydraulic System and RPS. 4. The geographical relationship between the full-core display and the East and West Scram Discharge Volumes (SDV). 5. Expected Rod Position indications for full-in rods and rods partially inserted between reed switches. Demonstrate the ability required by 1-EOI-1 step RC/Q-21 to assess plant conditions and detennine the appropriate EOI Appendicies required to insert control rods under various failure modes. A is incorrect based on Item 1 and 2 above. This distractor becomes plausible based on the EOI Appedicies which would be appropriate if RPS did not initiate a full scram due to SDV high level. B is incorrect based on Item 2 and 4. The RPS Scram Groups do not match geographical indications on the full-core display. In addition, the EOI Appendicies chosen would not address these conditions. This distractor is plausible if the candidate becomes confused regarding Items 1 and 3. C is correct. D is incorrect based on Items 3 and 4. Geographical indications on the full-core display are incorrect for the East SDV. In addition, the EOI Appendicies chosen would not address these conditions. This distractor is plausible if the candidate becomes confused regarding Items 3 and 5.

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( 6. SRO295038G2.2.22 OOllC/AfflGlffRM 3.7.2//295038G2.2.221ISROONLY/12118/2007 RMS In order to prevent high off-site release rates, several features were incorporated into the design of the Off-Gas system in addition to limitations imposed by Technical Specifications. Which ONE of the following describes the Tech Spec LCO limit of concem , the design feature used to maintain operation within that limit, and the basis for that design feature? A~ Maintain H2 concentration ::;'4%. SJAE low pressure isolation to maintain sufficient Off-gas dilution flow. B. Maintain H2 concentration ::;'4%. Injection of 02 at the Hydrogen Recombiner inlet to reduce H2 concentration below explosive limits. C. Maintain effluents::;, 10 curies. Hold-up volume size is sufficiently large to allow radioactive decay. D. Maintain effluents::;, 10 curies. stack Dilution Fans provide sufficient dilution flow prior to discharge from the stack. KIA Statement: 295038 High Off-site Release Rate 2.2.22 - Equipment Control Knowledge of limiting conditions for operations and safety limits. KIA Justification: This question satisfies the KIA statement by requiring the candidate to identify design features to reduce off-site release rates which are related to Tech Spec LCOs and their bases. References: TRM 3.7.2 Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases. 0610 NRC SRO Exam

REFERENCE PROVIDED: None Plausibility Analysis: ( In order to answer this question correctly, the candidate must: Demonstrate knowledge of Limiting Conditions for Operation related to high off-site release rates located in the Technical Requirements Manual (TRM) and their bases. Included in the basis discussion are the design features associated with maintenance of the LCO within limits, and is therefore, testable knowledge. A is correct. B is incorrect based on the basis and design feature. It is plausible because the design feature in the distractor is correct for Hydrogen Injection, but the basis for Hydrogen Injection is to reduce Intergranular Stress Corrosion Cracking, not off-site release rates. C is incorrect based on the design feature, which is correct, but the LCO which does not apply to that design feature. It is plausible because the HOld-up volume is to allow for decay of gaseous effluent, however the 10 curie limit imposed by TRM 3.7.1 is a liquid effluent limit. D is incorrect for reasons similar to distractor C above but based on a different design feature. This is plausible because Stack Dilution Fans are designed to dilute gaseous releases prior to discharge off-site via the Main Stack.

Airborne Effluents TR 3.7.2 TR 3.7 PLANT SYSTEMS ( TR 3.7.2 LC03.7.2 Airborne Effluents Whenever the SJAE is in service, the concentration of hydrogen in the offgas downstream of the recombiners shall be limited to ~ 4% by volume. APPLICABILITY: During main condenser offgas treatment system operation

NOTE--,

TRM LCO 3.0.3 is not applicable. ACTIONS

A. CONDITION With the concentration A.1 of hydrogen >4% by volume. REQUIRED ACTION Restore the concentration to within the limit. COMPLETION TIME 48 hours TECHNICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY TSR 3.7.2.1 BFN-UNIT 3 The concentration of hydrogen downstream of the recombiners shall be determined to be s 4% by volume by monitoring the off-gas whenever the SJAE is in service using instruments described in Technical Requirement 3.3.9. 3.7-3 Continuously by at least one OPERABLE Offgas Hydrogen Analyzer As required by TR 3.3.9 when Offgas Hydrogen Analyzer instrumentation is inoperable TRM Revision 0

TR 3.7 PLANT SYSTEMS ( TR 3.7.2 BASES Airborne Effluents B 3.7.2 Airborne Effluents BACKGROUND APPLICABLE SAFETY ANALYSIS The main condenser offgas system at BFN has a modification from the original installation by the addition of Recombiners which cause catalytic recombination of hydrogen and oxygen formed from the radiolytic decomposition of water in the reactor. After being recombined in the gas phase, the steam is cooled to cause condensation which is returned to the hotwell. The recombination causes a drastic reduction in offgas volume, enough to turn the 30-minute holdup volume into a 6-hour holdup volume. The extra retention time results in a reduction of offgas activity (curies). One of the power generation design basis of the offgas system is to provide adequate safeguards against possible explosion hazard of the hydrogen and oxygen present. In order to monitor for explosions and explosive conditions, the system has the following instrumentation that annunciate and/or are indicated in the control room: 1. Offgas flow 2. Temperature 3. Hydrogen Analyzers Establishment of limits for concentrations of hydrogen downstream of the offgas recombiners and a surveillance program to ensure the limits are maintained are required by the Explosive Gas and Storage Tank Radioactivity Monitoring Program of Technical Specification 5.5.8. The hydrogen concentration of the gases from the air ejector is maintained below the flammable limit (4%) by maintaining adequate steam flow for dilution at all times. The pressure of the steam supplied to the first and third stage steam jet air ejectors is monitored. The steam jet air ejector inlet and effluent are automatically isolated on low steam supply pressure. The preheaters are heated with steam, rather than electrically, to eliminate presence of potential ignition sources and to limit the BFN-UNIT 3 B 3.7-3 TRM Revision 0

( BASES APPLICABLE SAFETY ANALYSIS (continued) LCO 3.7.2 APPLICABILITY ACTIONS Airborne Effluents B 3.7.2 temperature of the gases in the event of cessation of gas flow. The recombiner temperatures are monitored and an alarm is actuated to indicate any deterioration of performance. A hydrogen analyzer downstream of the recombiners provides an additional check on recombiner performance. Maintaining the concentration of hydrogen below its flammability limit provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50. Four percent hydrogen concentration in air is the lower flammable limit for the mixture. The hydrogen buildup in the offgas system will stop when the main condenser offgas system is removed from service. Hence, this requirement is only applicable during main condenser offgas treatment system operation. Forty-eight hours is sufficient time to correct the conditions which increased the hydrogen concentration. If this time is exceeded, then conditions exist which could result in an uncontrolled hydrogen burn and removal of the offgas system from service will allow purging operations to decrease the hydrogen concentration. To stop the hydrogen production within the reactor vessel will require shutdown. Plant procedures are in place that require both shutdown of the reactor and removal of the main condenser offgas process. BFN-UNIT 3 B 3.7-4 TRM Revision 0

BASES TECHNICAL SURVEILLANCE REQUIREMENTS REFERENCES Airborne Effluents B3.7.2 TSR 3.7.2.1 Continuous monitoring by the offgas hydrogen analyzer alerts the operator of explosive conditions within the offgas system, and prompt the operator to take immediate actions to reduce the chance of a hydrogen explosion in the offgas system. If the offgas hydrogen analyzer instrumentation is inoperable, the ACTIONS of TRM LCO 3.3.9 ensure that frequent monitoring is continued. 1. BFN Technical Specifications (version prior to standardized version) 2. Section 9.5 of BFN FSAR BFN-UNIT 3 B 3.7-5 TRM Revision 0

Liquid Effluents TR 3.7.1 TR 3.7 PLANT SYSTEMS TR 3.7.1 Liquid Effluents LCO 3.7.1 The maximum activity to be contained in one liquid radwaste tank or temporary storage tank that can be discharged directly to the environs shall not exceed 10 curies excluding tritium and dissolved/entrained noble gas. APPLICABILITY: At all times

NOTE-------------------------------------------------------

TRM LCO 3.0.3 is not applicable. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Radioactive liquid A.1 Suspend all additions of Immediately waste exceeding the radioactive material to limits ofTRM LCO the tank. 3.7.1. AND A.2 Reduce the tank contents 48 hours to within the limit. AND A.3 Report events leading to Next Annual this condition in the next Radioactive Effluent Annual Radioactive Release Report Effluent Release Report (Section 5.2 of the ODCM). ( BFN-UNIT 3 3.7-1 TRM Revision 0

7. SRO 600000G2.2.24 OOl/CIAlSSI/EECW//600000G2.2.241ISRO ONLY/12118/2007 RMS During operation at 100% power on both units, Fire Pump "A" is tagged to trouble shoot problems encountered in the control circuit while testing. Which ONE of the following actions satisfy the Equivalent Shutdown Capability (ESC) for Fire Pump "A"? (Consider all other systems are operable.) A. Provide an engineering evaluation and a change to the Appendix R Safe Shutdown Program that provides an altemate method to perform the Appendix R function. B. Immediately obtain PORC and Plant Manager approval to utilize the diesel fire pump instead of Fire Pump "A". C." Provide a change to the SSls and to the safe shutdown program such that Fire Pump "A" is no longer required to comply with Appendix R requirements. D. Restore all Fire Pump HAlt functions to OPERABLE status in accordance with Appendix R SSI requirements in 60 days. KIA Statement: 600000 Plant Fire On-site 2.2.24 - Equipment Control Ability to analyze the affect of maintenance activities on LCO status. KIA Justification: This question satisfies the KIA statement by requiring the candidate to demonstrate knowledge of the affect of maintenance activities on Appendix R related components including LeO actions and compensatory measures required. References: Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of information. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions. 0610 NRC SRO EXam

REFERENCE PROVIDED: None Plausibility Analysis: ( In order to answer this question correctly, the candidate must: Recognize that in accordance with Fire Protection Report Volume 1, Appendix R Safe Shutdown Program, all functions are removed when the component is tagged. Determine which of the given conditions qualifies as an Equivalent Shutdown Capability as defined by the Fire Protection Report Volume 1, Appendix R Safe Shutdown Program. An Equivalent Shutdown Capability is limited to three possible options. The correct answer is Option 3. However, the distractors are plausible based on the actions required AFTER an Equivalent Shutdown Capability has been performed. This subtle difference has previously been a confusing issue at BFN and therefore, is an important skill required of licensed SROs to ensure compliance with Appendix R requirements. A is incorrect. This is performed only when an Equivalent Shutdown Capability has been used and the component has not been restored after 60 days. B is incorrect. This is performed only to ensure the adequacy of an Equivalent Shutdown Capability if the component has not been restored after 60 days. C is correct. D is incorrect. This action is only applicable AFTER an Equivalent Shutdown Capability has been established.

( Manual #: Fire Protection Report PLANT: BFN UNIT(s):1/2/3 PAGE 462 of915 Vol. 1 TITLE: Appendix R Safe Shutdown Program SEcrION: 4 REV: 0 The listed compensatory measure in the Unit I, 2 & 3 tables due to equipment degredation or the compensatory measures due to lack of spatial separation per 9.3.11.G.l.a of the Fire Protection Plan may be removed/revised if:

the affected unit is brought to COLD SHUTDOWN, or

an engineering analysis is performed, this program is changed and the Safe Shutdown Instructions are changed to provide an alternative shutdown path or

a different compensatory measure or combination of measures is established (e.g., additional administrative controls, operator briefings, temporary procedures, interim strategies, operator manual actions, temporary fire barriers, temporary detection or suppression systems). An engineering analysis of the alternative measure should incorporate risk insights regarding the location, quantity and type of combustible material in the fire area; the presence of ignition sources and their likelihood of occurrence; the automatic fire suppression and fire detection capability in the fire area; the manual suppression capability in the fire area; and the human error probability where applicable. (Reference 5) Compensatory Measure A will be documented and tracked in accordance with Attachment A of this instruction. COMPENSATORY MEASURES A. Restore the equipment function in 7 days or provide equivalent shutdown capability by one of the following methods. 1) A temporary alteration in accordance with plant procedures that allows the equipment to perform its intended function, or 2) A fire watch in accordance with the site impairment program in the affected areas/zones as specified in Section III. Note: Fire watch requirements in the Turbine Building (FA #25) and Control Building (FA #16) may be evaluated on a case by case basis due to the large size of these areas. For example, fire watches in the Turbine Building can be limited to within 20 feet of the south wall (near M-Line wall on EL 565' and 586') or the Intake Pumping Station due to the location of the RHRSW power cables in the areas. No Safe Shutdown circuits are located in any other location within the Turbine building. Control Building areas, even though not separated by fire resistive barriers, provide substantial protection against the spread of fire due to installed fire suppression systems and concrete floor slabs and walls. The potential of fire spread between control building compartments and the turbine building compartments has been evaluated in Section 3.0 of the IPEEE Fire Induced VUlnerability Evaluations for Units 1-3 (Reference A16 Section 3). These evaluations may be reviewed to determine the extent of fire watches.

Manual #: Fire Protection Report PLANT: BFN UNIT(s):1/2/3 PAGE 463 of915 Vol. 1 TITLE: Appendix R Safe Shutdown Program SECTION: 4 REV: 0 ( 3) A temporary change to the SSI's which provide safe shutdown without the required function. If equivalent shutdown capability is used, restore the equipment function in 60 days or provide an engineering evaluation and a change to this program that provides an alternate method to perform the Appendix R function, otherwise provide PORC review and Plant Manager Approval of the equivalent shutdown capability to ensure its adequacy. This review shall be conducted every 60 days until an alternate method is in place. Site Engineering may be contacted for assistance in determining what constitutes equivalent shutdown capability. An example would be the use of the spare SHDN BD battery charger in lieu of one of the permanent SHDN BD chargers (i.e., 0 -CHGA-248-0000A, B, C, D or 3-CHGA-248-0003EB) .

8. SRO 29500202.4.6 OOl/CIAlEOI/INTROI5/29500202.4.6/1SRO ONLYI11129/07 RMS EOI-1 Path RC/P retainment override states: IF THE MAIN CONDENSER CAN BE MADE AVAILABLE THEN STABIUZE RPV PRESS BELOW 1073 PSIG WITH THE MAIN TURB BYPASS VALVES. ( If the main condenser is currently unavailable, which ONE of the following actions should be taken? A. Wait at that step until the main condenser can be made available, then continue in RCIP using the Main Turb Bypass Valves for pressure control. B.'II Continue to the next step and address other systems to augment pressure control, and if the main condenser can be made available at a later time then that step should be re-addressed. C. Continue at the next step and address other systems to augment pressure control, and if the main condenser can be made available at a later time and another entry condition into EOI-1 occurs, then when RC/P is re-addressed, the main condenser can be used. D. Wait at that step until the main condenser can be made available, or until directed to take other actions specified by the TSC. KIA Statement: 295002 Loss of Main Condenser Vac I 3 2.4.6 - Emergency Procedures I Plan Knowledge symptom based EOP mitigation strategies KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the appropriate mitigation strategy to restore the main condenser to service whife executing Emergency Operating Instructions. References: EOIPM, 1-EOI-1 Flowchart Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (5) Assessment of facifity conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. 0610 NRC SRO Exam (

REFERENCE PROVIDED: None Plausibility Analysis: ( In order to answer this question correctly, the candidate must: Determine the requirements regarding Retainment Override steps to include the following: 1. The action should be performed if current conditions allow, othelWise, continue following the flowpath to the next step. 2. If conditions change such that the step can now be accomplished after the Retainment Override step has been addressed and passed, it is appropriate to return to that step and complete the actions. A is incorrect. Based on Item 1 above, it is not appropriate to wait. This answer is plausible since controlling pressure using Main Turb Bypass Valves would be appropriate if the Main COndenser was made available. B is correct. C is incorrect. Based on Item 2 above, another EOI entry condition is not required to perform the actions directed by a Retainment Override step. This answer is plausible because of the requirement to re-enter EOls upon new entry conditions. In addition, failure to re-enter EOls following additional entry conditions has been an on-going issue with novice operator candidates. D is incorrect. Based on reason 1 above, it is not appropriate to wait. This becomes plausible because the TSC staff has the authority to provide specific direction to the control room in an emergency, however, this is only allowed when procedure guidance is not available or incorrect. This is clearly NOT the case in this condition. It also requires initiating 10CFR 50.54 (x) and (y) requirements.

USER'S GUIDE FOR EMERGENCY OPERATING INSTRUCTIONS EOI PROGRAM MANUAL SECTION O-V1I1-A RETAINMENT OVERRIDE STEP DISCUSSION: 1.... 1 Symbol- RETAINMENT OVERRIDE STEP A rectangle with a thick blue border represents a Retainment Override Symbol. Retainment Overrides are the only graphics symbol that have a blue border, and have a top line stating "WHILE EXECUTING THIS PROCEDURE" or "WHILE EXECUTING STEPS ... (applicable steps listed)" followed by a "IF" column and a "THEN" column. The flowline always enters the top center and exits the bottom center ofthe symbol. Use A Retainment Override Step contains a conditional action that the operator must continuously check for applicability while executing the steps that follow. Unless the Retainment Override Step specifically identifies the steps in the procedure that it applies to, it is applicable to all steps following it in the flow path. Retainment Overrides specify a condition with the "IF" logic statement and provide the required action to be taken, ifthe condition occurs, with a "THEN" logic statement More than one conditional action may be displayed in anyone Retainment Override Step. Each is separated by a single blue line. The entire Retainment Override symbol is given one step sequence number in the procedure, regardless ofthe number of conditional actions within the symbol. The logic terms and their use are discussed in Section 3.4.F ofthis User's Guide. REVISION 4 PAGE 19 OF 52 SECTION O-V1I1-A

USER'S GUIDE FOR EMERGENCY OPERATING INSTRUCTIONS EOI PROGRAM MANUAL SECTION O-VlII-A CONTINGENT ACTION STEP DISCUSSION: 1 ......1 Symbol- CONTINGENT ACfION STEP A rectangle with a red border and triangular corners represents a Contingent Action Step symbol. It is the only red bordered symbol appearing on the flowcharts. The flowlines always enter the top center and exit the bottom center ofthe symbol. Use The Contingent Action Step is similar to the Retainment Override Step in the sense that both are a type of Conditional Action Step. The step always presents a condition with a "WHEN" logic statement and then provides the required action to be taken, ifthe condition occurs, with a "THEN" logic statement The difference between the Retainment Override Step and the Contingent Action Step, regarding their use, is that when a Contingent Action Step is entered the operator is required to stop and wait for the "WHENII conditional statement to be met before proceeding with the action stated by the "THEN" statement Logic terms and their use are discussed in Section 3.4.F ofthis User's Guide " REVISION 4 PAGE 21 OF 52 SECTION O-VlII-A

( 9. SRO 295009G2.4.31 OOl/CIAfTIG2IEOIl/295009G2.4.3l11SRO ONLYI Given the following plant conditions:

Unit 2 is operating at full power with RCIC suction aligned to the Suppression Pool due to a flange leak on the CS&S suction valve.

The Main Turbine trips due to an EHC logic failure.

The Bypass Valves do not open to control RPV pressure after the trip.

Some control rods fail to insert from the scram signal.

The UO notes the following conditions - Reactor Power 4% - Suppression Pool Temperature 142°F - Drywell Pressure 8 psig - Reactor Water Level -140 inches slowly rising - RCIC injecting ZSO V (2.M rJ'I/ Bel 2.A is k -.e.. ~ ~~t=:-c< =" ~ H~GI LOGIC PO'lJER FAittlRE alalill U - SLC injecting with Tank level at 70% - Appendix 8B in progress, MSIV Dp 150 psig Which ONE of the following describes the required operator action? REFERENCE PROVIDED A. When SLC tank level reaches ~% , exit C5 and control RPV water level per RCIL. , ~ -...... r--: - - - - - - -

B B.estoreJ~e..V water level to between +2 inches an~51 inches with-RCIC. __ + s:-/ It C~ Secure RCIC and maintain RPV water level between -180 inches and level to which it was:towered. D. Secure RCIC and inject with HPCI in MANUAL control. .s1/"'f<:-.-

( KIA Statement: 295009 Low Reactor Water Levell 2 2.4.31 - Emergency Procedures I Plan Knowledge of annunciators alarms and indications, and use of the response instructions KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the appropriate actions to control reactor water level while executing Emergency Operating Instructions. References: 2-EOI-1 Flowchart, 2-EOI-C5 Flowchart Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. 0610 NRC SRO Exam REFERENCE PROVIDED: 2-EOI-1 Flowchart, 2-EOI-C5 Flowchart Plausibility Analysis: In order to answer this question correctly, the candidate must determine the following: 1. RCIC operation should not continue due to elevated Suppression Pool temperatures. (Caution 6) 2. HPCI operation is not possible in AUTO or MANUAL with a logic power failure. (ARP) 3. SLC injection means RPV level MUST be controlled using 2-EOI-C5 until the reactor is shutdown with control rods. 4. SLC tank level of 67% is required to raise level to +2 to +51 inches. 5. SLC tank level below 43% allows POWER to be controlled by 2-AOI-100-1, Scram, but NOT RPV water level. A is incorrect. SLC tank level below 43% allows reactor power to be controlled outside of EOls, but not RPV level. This is plausible because a SLC tank level of 43% allows RPV level to be raised to the normal range controlled by 2-EOI-1, RC/L f1owpath, but the controlling procedure remains 2-EOI-C5 until rods are inserted. B is incorrect. RPV level cannot be raised until SLC tank level drops another 3% and RCIC should not be used for injection due to elevated SP temperature. (Caution 6) This is plausible because the SLC tank level of 67% was recently lowered due to Alternate Source Term constraints. A tank level of 70% would have been acceptable before the change. C is correct. D is incorrect. Securing RCIC is the appropriate action, but HPCI will not run in MANUAL mode with the HPCI LOGIC POWER FAILURE annunciator in alarm. This is plausible without control board indications of the controller being de-energized to verify HPCI availability in MANUAL.

CAUTION

3

ELEVATED SUPPR CHMBR PRESS MAY TRIP RCIC I #6 HPCI OR RCIC SUCTION TEMP ABOVE 140 "F L L RCIQ.1S EJIITRC.'O lIND ENTDAa!-1lJ).1 .REAClOR~ '--

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( 10. SRO 500000EA2.02 OOl/C/AffIG2/CONTIPRI/2/500000EA2.02/3.5/SRO ONLY/ll/29/07 RMS Given the following plant conditions:

At 50% power on Unit 2, a loss of RPS A occurs.

RPS A is placed on the transformer, and all systems restored to normal except Containment H2-02 analyzer A, which remains isolated . Which ONE of the following describes the Tech Spec implications? REFERENCE PROVIDED A. Restore the "A" H2 analyzer to OPERABLE within 72 hours. No ACTION is required as long as the "B" O2 analyzer can be aligned to the drywell or torus. B." No ACTION is required due to the "A" H2 Analyzer being INOP. No ACTION is required as long as the "B" O2 analyzer can be aligned to the drywell or torus. C. No ACTION is required due to the "ANH2 Analyzer being INOP. Begin altemate O2 sampling and analyze the results immediately. D. Restore the NA" H2 analyzer to OPERABLE within 72 hours. Begin altemate O2 sampling and analyze the results immediately. KIA Statement: 500000 High CTMT Hydrogen Conc. I 5 EA2.02 - Ability to determine and I or interpret the following as they apply to HIGH PRIMARY CONTAINMENT HYDROGEN CONCENTRATIONS: Oxygen monitoring system availability KIA Justification: This question satisfies the KiA statement by requiring the candidate to use specific plant conditions to determine the operability of the Hydrogen-Oxygen Monitoring system. References: Unit-2 Technical Requirements Manual Level of Knowledae Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict.an outcome. This requires mentally using this knowledge and its meaning to predict.the correct.outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases. 0610 NRC SRO Exam

REFERENCE PROVIDED: TRM Sections 3.3.11 and 3.6.2 Plausibility Analysis: ( In orelerto answer this question correctly, the candidate must: Determine the limiting LCO for a loss of one Hydrogen Monitor channel. Determine the limiting LCO for a loss of one Oxygen Monitor channel. Recognize that either H2 or O2 Analyzer can sample either the drywell or suppression chamber. A is incorrect. Only one Hydrogen analyzer is required to be OPERABLE. This is plausible because the LCO for the Oxygen analyzer is correct. B is correct. C is incorrect. Alternate sampling is only required if the Oxygen analyzer cannot be aligned to either the drywell or suppression chamber. This is plausible because the LCO for the Hydrogen analyzer is correct. D is incorrect. Only one Hydrogen analyzer is required to be OPERABLE. Alternate sampling is only required if the Oxygen analyzer cannot be aligned to either the drywell or suppression chamber.

Hydrogen Monitoring Instrumentation TR 3.3.11 TR 3.3 INSTRUMENTATION TR 3.3.11 Hydrogen Monitoring Instrumentation LCO 3.3.11 APPLICABI L1TY: ACTIONS One drywell and suppression chamber hydrogen analyzer shall be OPERABLE MODE 1 during the time period a. From 24 hours after THERMAL POWER is > 15% RTP following startup, to b. 24 hours prior to reducing THERMAL POWER to < 15% RTP prior to the next scheduled reactor shutdown. CONDITION REQUIRED ACTION COMPLETION TIME A. No drywell hydrogen A.1 Restore one drywell 7 days analyzer operable. hydrogen analyzer to OPERABLE status. B. No suppression B.1 Restore suppression 7 days chamber hydrogen chamber hydrogen analyzer operable. analyzer to OPERABLE status. C. Required Action and C.1 Initiate a Problem 24 hours associated Completion Evaluation Report Time of Condition A or (PER)/Corrective Actions B not met. Program (CAP) document to develop plans and schedule for restoring the analyzer to OPERABLE status. BFN-UNIT 2 3.3-61 TRM Revision 48, 53 October 21, 2005

TR 3.6 CONTAINMENT SYSTEMS ( TR 3.6.2 Oxygen Concentration Monitors TR 3.6.2 Oxygen Concentration Monitors LCO 3.6.2 APPLICABI L1TY: ACTIONS Primary Containment oxygen concentration monitors shall be OPERABLE. MODE 1 during the time period a. From 24 hours after THERMAL POWER is > 15% RTP following startup, to b. 24 hours prior to reducing THERMAL POWER to < 15% RTP prior to the next scheduled reactor shutdown. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more primary containment oxygen concentration monitors inoperable.

NOTE-----

Only applicable if there is loss of monitoring capability of the drywell or suppression chamber. A.1 Begin alternate sampling Immediately and analyze results. Every 7 days thereafter TECHNICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY TSR 3.6.2.1 BFN-UNIT 2 Perform CHANNEL CALIBRATION. 3.6-4 Every REFUELING OUTAGE TRM Revision 0

E MINATION REFERENCE ( PROVIDED TO CANDIDATE

Hydrogen Monitoring Instrumentation TR 3.3.11 TR 3.3 INSTRUMENTATION TR 3.3.11 Hydrogen Monitoring Instrumentation LCO 3.3.11 APPLICABIL1TY: ACTIONS One drywell and suppression chamber hydrogen analyzer shall be OPERABLE MODE 1 during the time period a. From 24 hours after THERMAL POWER is > 15% RTP following startup, to b. 24 hours prior to reducing THERMAL POWER to < 15% RTP prior to the next scheduled reactor shutdown. CONDITION REQUIRED ACTION COMPLETION TIME A. No drywell hydrogen A.1 Restore one drywell 7 days analyzer operable. hydrogen analyzer to OPERABLE status. B. No suppression B.1 Restore suppression 7 days chamber hydrogen chamber hydrogen analyzer operable. analyzer to OPERABLE status. C. Required Action and C.1 Initiate a Problem 24 hours associated Completion Evaluation Report Time of Condition A or (PER)/Corrective Actions B not met. Program (CAP) document to develop plans and schedule for restoring the analyzer to OPERABLE status. BFN-UNIT 2 3.3-61 TRM Revision 48, 53 October 21, 2005

TR 3.6 CONTAINMENT SYSTEMS ( TR 3.6.2 Oxygen Concentration Monitors TR 3.6.2 Oxygen Concentration Monitors LCO 3.6.2 APPLICABILITY: ACTIONS Primary Containment oxygen concentration monitors shall be OPERABLE. MODE 1 during the time period a. From 24 hours after THERMAL POWER is > 15% RTP following startup, to b. 24 hours prior to reducing THERMAL POWER to < 15% RTP prior to the next scheduled reactor shutdown. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more primary containment oxygen concentration monitors inoperable.

NOTE----*---

Only applicable if there is loss of monitoring capability of the drywell or suppression chamber. A.1 Begin alternate sampling and analyze results. Immediately AND Every 7 days thereafter TECHNICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY TSR 3.6.2.1 BFN-UNIT 2 Perform CHANNEL CALIBRATION. 3.6-4 Every REFUELING OUTAGE TRM Revision 0

( 11. SRO 203000A2.16 00 1IC/Aff2Gl/EOI C*1II203000A2.l61ISRO ONLY11211107 RMS Given the following plant conditions: ~as at 100% rated power near EOL. ~

A LOCA resulted in the following plant conditions: - All control rods are fully inserted. - RPV level is -120 inches and slowly lowering. ~ RPV pressure is 750 psig and slowly lowering. - Drywell pressure is 8.5 psig and slowly lowering. - RHR Loop I is aligned for Suppression Chamber Spray in accordance with 3-EOI Appendix 17C. - RHR Loop II is aligned for Drywell Spray in accordance with 3-EOI Appendix 17B. - Both loops of Core S ray are running on minimum flow with injection valves closed. - All three condensate p~ are running and available for injection. - 3-EOI C-1 f1owchartl.ffansitiOB::has just been performed.

=::---

~ 7 I Which ONE of the following describes the required procedure executions and the basis for that decision? REFERENCE PROVIDED A. - Direct 3-EOI Appendicies 1.6Fand 16G.,becomplete to enable RHR injection valve closure - Since two low pressure injection subsystems are available without RHR, Drywell and SP Sprays can be continued. - liljedion valve will REMAIN closed on a CAS signal but can be manually opened to allow injection if necessary. B. - D!~_e.t~E.QIAppendicies 1Qf~and ,t~ ,bELcompl~teJ()_eflab!~-.!!I:!R inje~ionyalve closure , _ / ::-Although two low pressure injection subsystems are available without RHR, Drywell and SP Sprays ) "--CANNOT be continued. ___ -Injection'valve will REMAINclosea on a-CAS*signaTb1.lrealfoenram:m1ly openedt<rallow- --'-- injection if necessary. C~ - Direct..3::EOI Appendicies.16E and16!:3 be complete to enable RHR injection valve.closure --- --- H lthough two low pressure injection subsystems areavaiiable'with-oUt RHR:Drywell and SP Sprays ") ~NOT be contmued. __--~-- _ ._..._~ - Injecllonvalve-will OPEN 'on s 'CASSignal but can be manually throttledtOcontror injaction-'once"- RPV level is above -162 inches and rising. D. -Direct LPC:~SY~ .L_~_<!.!LQ~T'§o.~L~~tVL~~S~ SEL in~switches ~ pla~dj~ BYP~ ~enWitfi two low pressure injection SUbsystemsavailable withoufRHR: Drywell and SP Sprays ANNOT be continued. ~__ _ _ - ./ - Injection vaive will OPEN on a CAS signal but can be manually ffiiOttled to-control injeCtioiiOiiCe RPV level is above -162 inches and rising.

( KIA Statement: 203000 RHRlLPCI: Injection Mode A2.16 - Ability to (a) predict the impacts of the following on the RHRlLPCI: INJECTION MODE (PLANT SPECIFIC) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Lo~ of coolant accident KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the appropriate procedures and their bases to control RHR in LPCI mode following a Loss of Coolant Accident. 6e~erencesu-EOI-1 Flowchart, 3-EOI-C1 Flowchart, EOIPM Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. 0610 NRC SRO Exam '7 REFERENCE PROVIDED: 3-EOI-1 and 3-EOI-C1 flowcharts Plausibility Analysis: In order to answer this question correctly, the candidate must determine the following: 1. EOI Appendix 16F and 16G are required to allow RHR LPCI injection control. 2. EOI Appendix 16F and 16G only apply to Unit-1 and Unit-3. 3. EOI-C1 requires LP systems lined up for injection regardless of containment control actions being performed at this point of EOI execution. 4. Although two or more systems lined up for injection is the ultimate goal of EOI-C1, all available injection systems should be lined up for injection if possible. 5. RHR LPCI injection valve operation with a CAS signal and RHR LPCI Injection Valve 5 minute timers bypassed. A is incorrect. Drywell and SP Sprays must be secured to line up RHR for injection. The RHR injection valve will still open on a CAS signal, but can be manually throttled close immediately if EOI Appendix 16F and 16G have been completed. This is plausible since the correct EOI Appendix execution is part of the answer. In addition, failure to recognize the requirement to secure containment control actions has been a recurring issue with novice operators at BFN. B is incorrect. The RHR injection valve will still open on a CAS signal, but can be manually throttled close immediately if EOI Appendix 16F and 16G have been completed. This is plausible since part 1 and 3 of the answer are correct. C is correct. D is incorrect. LPCI SYS I and II OUTBD INJ VLV BYPASS SEL in switches are only installed on Unit-2. This is plausible because this modification was scheduled to be installed on Unit-3 but has been postponed due to budget constraints.

( + P FORM TlIE FO'.lOWNG WITH USTEO INJ SUBSYSTEMS IRRESPECTIVE OF PUMP NPSH AND VORTEX UMJTS: 1. LIN UP FORINJ 2. TARTPU P S 3. RAISE INJflOW10 TIE .WI( INJSUBSYSTEM !'PPX L'lJPRESS CNDS AND FW SA 1210 PSlG CRn 59 1640 PSlG RClC. WITH CST SUCTION IF POSSIBLE 5C 1240PSIG HPC\\ WITH CST SUC ON IF POSSIBlE SO 1240PSIG CNOS 6A 410PSIG LPCI SYSTEM I (pUM?S A OR C) 68 320PSIG LPCI SYSTEM II (PU IPS B OR D) 6C 320PSIG CS SYSTEM I (Pln fPS A OR C) 60 330PSlG CSSYSTEM II (puMPS B OR D) lie 330 PSlG SYSTEMS lISTED IN SIEPC1-'>.IF - REQUiRm TO AUGMENT NJECIlON - L C1-4 ~, CMI 2 0R .toRE NO L CNDS, LPCI, OR CS ... INJSU3SYSTEMSBE ... "\\ AI..IGNED WllH PU IPS RlJNN G C1-6 vesL ,Ir

( 3-EOI APPENDIX-6B Rev. 02

.. pa.ge 1 of 3 3-EOI APPENDIX-6B INJECTION SUBSYSTEMS LINEUP RHR SYSTEM I LPCI MODE LOCAT:ON: Unit 3 Control Room ATTACHMENTS: 1 . NPSH Monitoring 1. IF ..... Adequate core cooling is assured AND I~ becomes necessary to bypass LPCI Injection Valve Timers to control injection, THEN ... EXECUTE EOI Appendix 16F concurrently with this procedure . 3-EOI APPENDIX-6C Rev. 02 .. iiPiage 1 of 3 3-EOI APPENDIX-6C INJECTION SUBSYSTEMS LINEUP RHR SYSTEM II LPCI MODE LOCAT:ON: Unit 3 Control Room ( A~TACHMENTS: 1. NPSH Monitoring 1. IF ..... Adequate core cooling is assured Ie becomes necessary to bypass LPCl ~njection Valve Timers to control injection, THEN ... EXECUTE EOl Appendix 16G concurrently with this procedure. (v)

( This appendix provides instruction on venting of the RPV when necessary to allow for containment flooding. If any gross fuel failure exists, Offgas release rate limits may be exceeded when . implementing this procedure. Step C1-32 directs venting of the RPV per Appendix 15 when Primary Containment Water Level reaches 30 ft. 16. Appendices 16A-16L These appendices provide instruction for bypassing RCIC, HPCI, and/or RHR interlocks as necessary based upon plant conditions. a. Appendices within this category are as follows: 16A Bypassing RCIC Low Reactor Pressure Isolation Interlocks 16B Bypassing RCIC Test Mode Isolation Interlocks 16C Bypassing HPCI Low Reactor Pressure Isolation Interlocks 160 Bypassing HPCI Test Mode Isolation Interlocks 16E Bypassing HPCI High Suppression Pool Water Level Suction Transfer Interlock 16F Bypassing the Loop I LPCI Injection Valve Timer 16G Bypassing the Loop II LPCllnjection Valve Timer 16H Bypassing RCIC High RPV Level Isolation Interlocks OPL171.206 Revision 7 Appendix 1 Page 27 of 42 release rates of Table 7 apply while venting NLO NLO NLO NLO NLO Not used on Unit 2 Not used on Unit 2 NLO

EOI PROGRAIiIlANUAL SECTlON o-Y-o STLP: CI--t C1. ALTERNATE LEVEL CONTROL BASES .

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SEcnON o-V-G PAGE 14 OF 50 REVISION 0

EOI PROGRAM MANUAL SECllON o-V-G . DISCUSSION: STEP Cl-4 C1, ALTERNATE LEVELCONTROL BASES 1 1 This action step directstheoperatorto take manualcontrolofat least two injectionsubsystems to provideassurancethat water willbe injected intothe RPVduringand followingRPV depressurization. A requirementfor at leasttwo injectionsubsystemsis specifiedinthis step to accommodate thepossibilitythat one subsystemmaynot operateproperly,or thata break. may existin the flowpathofoneinjectionsubsystem. EOIAppendicesprovidestep-by-step guidance for injectingintotheRPV. Injectionpressures(<AI>> havealso beenprovided as additional informationfor theoperator. Unlikedirectionsgivenfor use ofmotordrivenECCSpumpsin the RCILSectionofEaI-I, RPV Control,operationofECCSpumps in this procedureis carriedout irrespective ofassociated NPSHand Vortexlimits(suppressionpool level). Eventhoughrisk ofequipmentdamageexistsif NPSH andVortexlimits are exceeded,immediateandcatastrophicpump failureis not expected shouldoperationbeyondthese limitsbe required. Sinceprolongedoperationunderthese conditionsis most likelyrequired beforedegradedsystemandpumpperformance mayresult,the undesirable consequences ofuncoveringthe reactorcore outweighthe risk ofequipmentdamage. Byraisinginjectionflowto maximum,maximumflowwill be delivered to the RPVas soonas RPVpressuredrops below shutoffhead pressureofoperatinginjectionsubsystempumps. This actionpromotesrapid recoveryofRPV water level. Subsequentactions will controlinjectionand maintainRPV water level in the desiredband. Listedinjectionsubsystemsare limitedto thosebavingmotor-drivenpumps. Steam-driven systemsare notclassifiedasinjectionsubsystems becausetheymaynot be availablefollowing RPVdepressurization. Injectionsubsystemsaredefinedby physicalseparationofcomponents, flowpaths. and injectionpoints. An injectionsubsystem,as identifiedin this step, is a motor- drivensystem loopwhich is independently capableofsupplyingmakeupwater to the RPV. To illustrate:

RHR System. is comprisedoftwo LPCIsubsystems, each consistingoftwo pumpswith independentsuctionand dischargeflowpathsintothe RPV.

CondensateSystemis comprisedofthree hotwellpumps and three condensatebooster pumps whichdischargeinto a commonheader. This is oneinjectionsubsystem, not three.

CS Systemis comprisedoftwo CS subsystems, eachconsistingoftwo pumpswith independentsuctionand dischargeflowpathsintotheRPV. " REVISION 0 PAGE 150FSO SECTlON O-V-G

E MINATION REFERENCE PROVIDED TO CANDIDATE

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A. ( ( 12. SRO 21500302.1.14 OOllC/A/GOIl00-IAIIRMIB10/21500302.1.14//SRO ONLYI1211107 RMS Given the following plant conditions:

Unit-3 is perfonning a startup and heatup in accordance with 3-GOI-100-1A, UNIT STARTUP.

All IRMs are OPERABLE on Range 7 when a half-scram trip occurs on RPS "A".

The Reactivity Manager reports the cause of the half-scram was due to a momentary upscale trip on IRM "G", but the IRM is currently reading nonnally.

No IRM Range Switches were being manipulated at the time the half-scram trip was received. Which ONE of the following describes the required actions? REFERENCE PROVIDED --_.._--- ~ ~e control rod withdrawal nd obtain Plant Manager's pennission to place IRM "G" in bypass. - eset the half-scram and continue with the heatup. - Enter LCO 3.3.1.1.A1 on IRM "G". B. -..:: St~~tml...rod:~nd obtain Plan~nager's pennission to place IRM "G" in bypass. - Reset the half-scram and monitor IRM "G" for 15 minutes. - Enter an INFORMATION ONLY LCO on IRM "G".

-- ----.,

C." cSfo~ntrol rod withdral, place IRM "G" in bypass and notify the System Engineer. - Reset the half-scram and monitor IRM "G" for 15 minutes. ~ - Enter an INFORMATION ONLY LCO on IRM "G". . . .~- ----- ----- D. 0Stop control rod withdra~place IRM "G" in bypass and notify the System ~neer. --Reset the.haJf-seranrana monitor IRM "G" for 15 minutes. - . - Enter LCO 3.3.1.1.A1 on IRM "G". KIA Statement: 2150031RM 2.1.14 - Conduct of Operations Knowledge of system status criteria which require the notification of plant personnel KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to detennine the appropriate actions and notifications due to an IRM failure. References: 3-01-92A, 3-ARP-9-5A (33), OPDP-8, TS 3.3.1.1 Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases. 0610 NRC SRO Exam

REFERENCE PROVIDED: U3 Tech Spec Section 3.3.1.1 Plausibility Analysis: ( In order to answer this question correctly, the candidate must determine the following: 1. 3-ARP-9-5A (33) directs that IRM "G" be placed in BYPASS in accordance with 3-01-92A. 2. Plant Manager's permission is NOT required since the bypass is directed by an approved procedure. 3. 3-ARP-9-5A (33) directs that the scram be reset. 4. 3-01-92A directs that the IRM be monitored for 15 minutes following a spike which results in a scram trip. 5. System Engineering must be notified to obtain concurrence prior to returning the IRM to service. 6. Tech Spec 3.3.1.1 requires 3 OPERABLE channels per trip system. IRM "G" is not required to be OPERABLE since all other IRMs assigned to trip system "A" are OPERABLE. 7. OPDP-8 directs an INFORMATION ONLY LCO be entered in the LCO Tracking Log. A is incorrect. Plant Manager's permission is not required to bypass IRM "G" and LCO 3.3.1.1.A1 is not appropriate. In addition, continuing with the heatup is not appropriate until the IRM is monitored for 15 minutes. This is plausible based on the subtle difference between an IRM that is "noisy" as opposed to an IRM that "spikes". B is incorrect. Plant Manager's permission is not required to bypass IRM "G H

This is plausible since

the second and third part of the answer are correct. C is correct. D is incorrect. LeO 3.3.1.1.A1 is not appropriate. This is plausible since the first and second part of the answer are correct.

BFN Intennediate Range Monitors 3-01-92A Unit 3 Rev. 0014 Page 8 of 15 ( 3.0 PRECAUTIONS AND LIMITATIONS (continued) L. [IUFj An IRM or SRM may be bypassed in the following conditions: 1. STOP control rod withdrawal and PLACE the channel in bypass when the SRM or IRM first gets noisy. 2. STOP control rod withdrawal and PLACE the channel in bypass immediately upon receipt of a single event large noise spike. These conditions bypass the instrument for an operability assessment based on whether the noise is transitory or sustained. Transitory noise is considered a one time occurrence that does not repeat itself and the channel can be removed from bypass and restored to service. Sustained noise is when the duration exceeds 15 minutes and may result in signal build up until a trip signal is reached. If a trip or high flux signal was generated, the channel is required to be observed for at least 15 minutes before returning the instrument to service with concurrence from System Engineering. When the initial assessment and recognition of the magnitude of the event has been determined, then control rod withdrawal may be resumed where it has been left off as long as the minimum number of SRM and IRM channels operable are within the Technical Specification limits. [1I-B-91-040] M. [ONe] SPP-10.4 requires approval of the Plant Manager or his designee prior to any planned operation with IRMs bypassed unless bypassing is specifically allowed within approved procedures. PSE-NPS-92-R01)

BFN----r-------------Pan;.-9-5- ---

-- - 13-ARP~9-5A ----------1

I ! Unit 3 3-XA-55-5A IRev. 0037 i _____________L -.J~C!Q!_~~r~ J ( IRM CHA, C, E, G HI-HI/INOP RED BAR f3:3 (Page 1 of 1) SensorfTrip Point: Relay K-16 A. HI-HI ~ 116.4 on 125 scale B.INOP. 1. Hi voltage low. 2. Module unplugged. 3. Function switch NOT in OPERATE. 4. Loss of+/- 24 VDC to monitor Sensor Location: Probable Cause: Automatic Action: Control Room Panel 3-9-12. A. Flux level at or above setpoint. B. One or more inoperable conditions exist. C. SI or SR in progress. D. Malfunction of sensor. E. Control rod drop accident. A. Half-scram if one sensor actuates (except with Rx Mode Sw. in RUN). B. Reactor scram if one sensor per channel actuates, (except with Rx Mode Sw. in RUN). operator Action: A. STOP any reactivity changes. B. VERIFY alarm by multiple indications. C. RANGE initiating channel or BYPASS initiating channel. REFER TO 3-01-92A.-... D. With SRO permission, RESET Half Scram. REFER TO 3-01-99 E. IF alarm is from a control rod drop, THEN REFER TO 3-AOI-85-1. F. [NRC/C) IF one or more IRM recorder reading is downscale, THEN CHECK for loss of +/- 24 VDC power. G. NOTIFY Instrument Maintenance that functional tests of any monitors indicating an INOP condition, including a downscale reading, are required before the instrument can be considered operable. [NRC IE item 86-40-{)3] H. NOTIFY Reactor Engineer. I. REFER TO Tech Spec Table 3.3.1.1-1, TRM Tables 3.3.4-1 and 3.3.5-1. oo o o o o o o References: 3-45E620-6 3-730E915RF-12 Technical Specifications 197R114-16 GEK 3-730E915-10 3-01-92A 3-AOI-85-1 Technical Requirements Manual-TRM

( 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. Place channel in trip. CONDITION A. One or more required channels inoperable. REQUIRED ACTION A.1 OR A.2

NOTE----------

Not applicable for Functions 2.a, 2.b, 2.c, 2.d, or 2.f. Place associated trip system in trip. COMPLETION TIME 12 hours 12 hours (continued) BFN-UNIT 3 3.3-1 Amendment No. 212, 213, 221 September 27. 1999

( Table 3.3.1.1-1 (page 1 of 3) Reactor Protection System Instrumentation RPS Instrumentation 3.3.1.1 FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS PER TRIP SYSTEM CONDITIONS REFERENCED FROM REQUIRED ACTION 0.1 SURVEILLANCE ALLOWABLE REQUIREMENTS VALUE 1. Intermediate Range Monitors a. Neutron Flux - High 2 3 G SR 3.3.1.1.1

s; 1201125

SR 3.3.1.1.3 divisions of full SR 3.3.1.1.5 scale SR 3.3.1.1.6 SR 3.3.1.1.9 SR 3.3.1.1.14 5(a) 3 H SR 3.3.1.1.1 s 120/125 SR 3.3.1.1.4 divisions of full SR 3.3.1.1.9 scale SR 3.3.1.1.14 b. lnap 2 3 G SR 3.3.1.1.3 NA SR 3.3.1.1.14 5(a) 3 H SR 3.3.1.1.4 NA SR 3.3.1.1.14 2. Average Power Range Monitors a. Neutron Rux - High, 2 3(b) G SR 3.3.1.1.1

s; 15% RTP

(Setdown) SR 3.3.1.1.6 SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16 b. Row Biased Simulated 3(b) F SR 3.3.1.1.1

s; O.66 W

Thermal Power - High SR 3.3.1.1.2 +66% RTP SR 3.3.1.1.7 and :s; 120% SR 3.3.1.1.13 RTP(c) SR 3.3.1.1.16 c. Neutron Aux - High 3(b) F SR 3.3.1.1.1

s; 120% RTP

SR 3.3.1.1.2 SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16 (continued) (a) IMth any control rod withdrawn from a core ceO containing one or more fuel assemblies. (b) Each APRM channel provides inputs to both trip systems. (c) [.66 W + 66% - .66 l! WJ RTP when reset for single loop operation per LCO 3.4.1, "Recirculation Loops Operating." BFN-UNIT3 3.3-7 Amendment No. 216 December 23, 1998

TVAN STANDARD DEPARTMENT PROCEDURE ( 3.3 OPDP-8 LIMITING CONDITIONS FOR OPERATION TRACKING Rev. 1 Page 5 of 24 TS LCO Evaluations 3.3.1 General Guidelines A. When equipment identified in TS is made or becomes inoperable, existing plant/unit conditions may require LCOs be entered. B. LCOs are entered if, for existing plant/unit conditions, TS require action(s) to be taken. C. LCOs are exited when the equipment is returned to operable status or when the plant/unit is put into a condition where TS no longer require action(s) to be taken. TS 3.0.6 and 5.7.2.18 (WBN)I5.5.11 (BFN) addresses LCO entry relative to support s~tems and supported systems. D. TS action requirements may change as the aggregate of inoperable equipment changes. Determination of TS action(s) and the quantity of LCOs to be entered are based on the aggregate of inoperable systems, equipment, and components. E. Multiple LCOs shall be entered and logged if equipment is inoperable or removed from service and more than one TS LCO action is required to be taken. F. If equipment is identified or made inoperable that does not apply to an LCO based on the current plant conditions, an "Information Only* LCO should be entered into the Unit Log or LCO Tracking Log, as appropriate. The "Information Only" LCO entry should contain information similar to an "Active" LCO with possibly the exception of the LCO expiration date not being required. During plant shutdown/outages, it is not required to utilize INFORMATION ONLY LCOs for conditions that are applicable only in other modes which are controlled by other plant instructions (i.e., general operating instructions, surveillance instructions etc.). 3.3.2 Tracking of Inoperable Eguipment A. When equipment identified in TS is discovered or determined to be inoperable, LCO action is assessed by the affected unifs US. This assessment is based on existing and planned near-term plant/unit conditions and evolutions. The assessment shall be in accordance with Appendix B (WBN) or Appendix C (BFN) "Safety Function Determination Program (SFDP)" in accordance with the WBN T5 5.7.2.18 orBFN TS 5.5.11. B. The affected US{s) shall identify ifthe unit requires an entry into an LCO based on the current plant conditions. c. C. If TS require that actions be taken, LCOs associated with the inoperable equipment shall be entered into the Unit Log and/or LCO Tracking Log{s) maintained by the affected unifs US/designee .

/ E MINATION REFERENCE (,PROVIDED TO CANDIDATE *

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. Place channel in trip. CONDITION A. One or more required channels inoperable. REQUIRED ACTION A.1 OR A.2 NOTE--- Not applicable for Functions 2.a, 2.b, 2.c, 2.d, or 2.f. Place associated trip system in trip. COMPLETION TIME 12 hours 12 hours (continued) BFN-UNIT 3 3.3-1 Amendment No. 212, 213, 221 September 27, 1999

RPS Instrumentation 3.3.1.1 CTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. -------------NOTE------------- B.1 Place channel in one trip 6 hours Not applicable for system in trip. Functions 2.a, 2.b, 2.c, 2.d, or 2.1. OR

8.2 Place one trip system in 6 hours One or more Functions trip. with one or more required channels inoperable in both trip systems. C. One or more Functions C.1 Restore RPS trip 1 hour with RPS trip capability capability. not maintained. D. Required Action and 0 .1 Enter the Condition Immediately associated Completion referenced in Time of Condition A, B, or Table 3.3.1.1-1 for the C not met. channel. E. As required by Required E.1 Reduce THERMAL 4 hours Action 0 .1 and POWER to < 30% RTP. referenced in Table 3.3.1 .1-1. F. As required by Required F.1 Be in MODE 2. 6 hours Action 0.1 and referenced in Table 3.3.1.1-1. A ( (continued) BFN-UNIT3 3.3-2 Amendment No. 212, 213, 221 September 27, 1999

RPS Instrumentation 3.3.1.1 CTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME G. As required by Required G.1 Be in MODE 3. 12 hours Action D.1 and referenced in Table 3.3.1.1-1 . H. As required by Required H.1 Initiate action to fully Immediately Action D.1 and insert all insertable referenced in control rods in core cells Table 3.3.1.1-1. containing one or more fuel assemblies. I. As required by Required 1.1 Initiate alternate method 12 hours Action D.1 and to detect and suppress referenced in Table thermal hydraulic 3.3.1.1-1. instability oscillations. J. Required Action and J.1 Be in MODE 2 4 hours associated Completion Time of Condition I not met. A ( BFN-UNIT 3 3.3-3 Amendment No. 212, 213, 221, 231 September 13,2001

( RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS

NC>TES----------------------------------------------------

1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function. 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains RPS trip capability. SR 3.3.1.1.1 SR 3.3.1.1.2 SR 3.3.1.1.3 BFN-UNIT 3 SURVEILLANCE Perform CHANNEL CHECK.

NC>TE-------------------------

Not required to be performed until 12 hours after THERMAL PC>WER ~ 25% RTP. Verify the absolute difference between the average power range monitor (APRM) channels and the calculated power is s 2% RTP while operating at ~ 25% RTP.

NC>TE

Not required to be performed when entering MC>OE 2 from MC>OE 1 until 12 hours after entering MC>OE 2. Perform CHANNEL FUNCTIC>NAL TEST. 3.3-4 FREQUENCY 24 hours 7 days 7 days (continued) Amendment No. 213 September 03,1998

( RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.1.1.4 Perform CHANNEL FUNCTIONAL TEST. 7 days SR 3.3.1.1.5 Verify the source range monitor (SRM) and Prior to intermediate range monitor (IRM) channels withdrawing overlap. SRMs from the fully inserted position SR 3.3.1.1.6

NOTE

-- Only required to be met during entry into MODE 2 from MODE 1.

Verify the IRM and APRM channels overlap. 7 days SR 3.3.1.1.7 Calibrate the local power range monitors. 1000 MWDfT average core exposure SR 3.3.1.1.8 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.1.1.9

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

1. Neutron detectors are excluded. 2. For Function 1, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.

Perform CHANNEL CALIBRATION. 92 days (continued) BFN-UNIT3 3.3-5 Amendment No. 213 September 03, 1998

( RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.1.1.10 Perform CHANNEL CALIBRATION. 184 days SR 3.3.1.1.11 (Deleted) SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST. 24 months SR 3.3.1.1.13

NOTE

Neutron detectors are excluded.

Perform CHANNEL CALIBRATION. 24 months SR 3.3.1.1.14 Perform LOGIC SYSTEM FUNCTIONAL 24 months TEST. SR 3.3.1.1.15 Verify Turbine Stop Valve - Closure and 24 months Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions are not bypassed when THERMAL POWER is ~ 30% RTP. SR 3.3.1.1.16

NOTE-------------------------

For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.

Perform CHANNEL FUNCTIONAL TEST. 184 days SR 3.3.1.1.17 Verify OPRM is not bypassed when APRM 24 months Simulated Thermal Power is ~ 25% and recirculation drive flow is < 60% of rated recirculation drive flow. BFN-UNIT 3 3.3-6 Amendment No. 212, 213, 215, 221 September 27, 1999

RPS Instrumentation 3.3.1.1 ( ~ Table 3.3.1.1-1 (page 1 of 3) Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VAlUE CONDITIONS SYSTEM ACTION 0 .1 1. Intermediate Range Monitors a. Neutron Flux - High 2 3 G SR 3.3.1.1.1 s 120/125 SR 3.3.1.1.3 divisions of full SR 3.3.1.1.5 scale SR 3.3.1.1.6 SR 3.3.1.1.9 SR 3.3.1.1.14 5(a) 3 H SR 3.3.1.1.1

s120/125

SR 3.3.1.1.4 divisions of full 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 5(a) 3 H SR 3.3.1.1.4 NA SR 3.3.1.1.14 2. Average Power Range Monitors a. Neutron Aux - High, 2 3(b) G SR 3.3.1.1.1 s 15% RTP (Setdown) SR 3.3.1.1.6 SR 3.3.1.1 .7 SR 3.3.1 .1.13 SR 3.3.1.1.16 b. Aow Biased Simulated 3(b) F SR 3.3.1.1.1

s0.66 W

Thermal Power - High SR 3.3.1.1.2 +66%RTP SR 3.3.1.1.7 and:s 120% SR 3.3.1.1.13 RTP(c) SR 3.3.1.1.16 c. Neutron Flux - High 3(b) F SR 3.3.1.1.1

s120% RTP

SR 3.3.1.1.2 SR 3.3.1 .1.7 SR 3.3.1.1.13 SR 3.3.1.1.16 (continued) (a) Withany control rod withdrawn from a core cell containing one or more fuel assemblies. (b) Each APRM channel provides inputs to both trip systems. (c) [.66 W + 66% - .66 ~ WJ RTP Ylhen reset for single loop operation per LCO 3.4.1, "Recirculation Loops Operating." c,. BFN-UNIT 3 3.3-7 Amendment No. 216 December 23, 1998

( 13. SRO 259002G2.1.23 OOllC/Aff2Gl/CAUTION1I3/259002G2.1.23/4.0/SRO ONLY/12/1107 RMS Given the following Unit 2 conditions:

Reactor pressure: 10 psig

Drywell temperature: 250°F

Secondary Containment temperatures 74-95f 74-95c &d 69-835a thru d 69-29f, g & h 220°F 245°F 260°F 200°F

Reactor water level indications: L1-3-58a & b L1-3-52 & 62a L1-3-53, 60 & 206 L1-3-55 Erratic -150 inches oinches oinches ( Which ONE of the following describes the required action and the basis for that action? REFERENCE PROVIDED A. Enter 2-EOI-C-4, RPV Flooding due to non-condensible gases coming out of solution in the reference legs of L1-3-58a & b as a result of low reactor pressure. B. Enter 2-EOI-C-1, Alternate Level Control due to non-condensible gases coming out of solution in the reference legs of L1-3-58a & b as a result of low reactor pressure. C. ttl Enter 2-EOI-C-4, RPV Flooding due to flashing steam in the reference legs of L1-3-58a & b as a result of low reactor pressure. D. Enter 2-EOI-C-1, Alternate Level Control due to flashing steam in the reference legs of L1-3-58a & b as a result of low reactor pressure. KIA Statement: 259002 Reactor Water Level Control 2.1.23 - Conduct of Operations Ability to perform specific system and integrated plant procedures during all modes of plant operation KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the appropriate procedure and its bases to control reactor water level. References: 2-EOI-C-4 Flowchart, EOIPM Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. 0610 NRC SRO Exam

REFERENCE PROVIDED: EOI Caution 1, Curve 8, Table 6, PIP 95-64 Plausibility Analysis: ( In order to answer this question correctly, the candidate must determine the following: 1. RPV pressure vs. temperature is in the UNSAFE area of Curve 8, RPV Saturation Temp. 2. Perform corrections to L1-3-52& 62a using PIP 95-64 to determine their proximity to TAF. 3. Recognize L13-53, 60, 206 & 55 are indicating below their Minimum Indicated Level for the current temperatures in the Secondary Containment. 4. Recognize that erratic indication on L1-3-58a & 58b is indicative of reference leg flashing. 5. Recognize the difference between "flashing" and "notching" with respect to indicated instrument response. A is incorrect. Non-condensible gases are indicative of "Notching" instruments caused by rapid depressurization. This is plausible because the correct EOI flowchart is entered. Recent issues at BFN indicate the propensity for novice operators to confuse "Notching" and "Flashing" conditions. B is incorrect. Non-condensible gases are indicative of "Notching" instruments caused by rapid depressurization. In addition, the incorrect EOI flowchart is entered. This is plausible because L1-3-52 and 62a are approaching TAF, which would indicate a need to transition to 3-EOI-C1, Alternate Level Control. C is correct. D is incorrect. Flashing steam in the reference legs of L1-3-58a & b is indication of a loss of RPV level instrumentation since Curve 8, RPV saturation Temp is currently unsafe. However, the correct action is to enter 3-EOI-C4, RPV Flooding. This is plausible because L1-3-52and 62a are approaching TAF, which would indicate a need to transition to 3-EOI-C1. Alternate Level Control.

( CAUTIONS CAUTION #1

AN RPV WATERlVL INS

MENTMAYBE USB:> 0 DETERMI E ORTRENDLVI..0 Y 'II EN I READSABOVE THE I I M OICATB:>LVL ASSOCIA 0\\1'/1 E IGHES MAXOWOR BeR TEMP.

11= ow TEMPS. OR se AREA EMPS(TABlE 8~ AS APPUCABLE.ARE0 IDETHESAFE REGIO OF CURVE8. EASSOCIA 0 I S WE MAYSE REUA8lE OUETOB01U GI ER N. MINIMUM MAX OW RUN TEMP MAXSC INSTRUMENT RANGE INDICATED (FROM XR-64-50 RUN TEMP LVL OR TI-64-52AB) (FROM TABLE 6) ON SCALE N/A BELOW 150

145

N/A 151 T0 200 LI-3-58A..B EMERGENCY -140 N/A 201 TO 250 -155 TO +60

130

N/A 251 ro soo

120

N/A 301 10 350 LJ.3.53 ON SCALE N/A BELOW 150 LI-3-6O +5 NlA 151 TO 200 LJ..3.200 NORMAl.. +15 N/A 20110250 OTO+60 Ll-3-253 +20 N/A 25110 soo U-3-206A, B,C. 0 +30 N/A 301 10350 ll-3-52 POST L\\.3-62A ACCIDENT ON SCALE N/A NlA -268 TO +32 +10 BELOW 100 NlA +15 100 TO 150 NlA SHUTDOWN +:!I 151 T02CO NlA 1I-3-55 FlOOOUP +30 201 TO 250 NlA oTO +40{) +40 251 T0 3CO NfA +50 301 TO 350 NlA +65 351 T0 4()O NJA

CURVE 8 RPV SATURATIO TEMP -- ".,....- -- ACTION ~ IREQUlREO I/' /' / I SAFE r / I t / l 400 380 t.360 co ~ 3 ffi 320

!i: I:80

! 260

il:

m 240 220 200 o 50 1 150 RPV PRESS (ps1G)

200 250 ....CONSTANT ABOVE 200 PSlG TABLES SECONDARY CONTMT INSTRUMENT RUNS INSTRUMENT SC TEMP ELEMENTS AND LO CATIONS EL621 EL593 EL565 RWCUHXRM (74-95F) (74-95C AND 0) (69--83&\\ THRU 0) (69-29F. G , H) U-3-58A "F "F NlA OF U-3-58B e tF NlA N/A 1I-3-53 it IlF NfA OF U-3-60 c Cf NtA N/A L1-3-206 !IF "F NlA OF 1I-3-253 IT Cf NIA N/A L1-3-52 CF Cf cF N/A U-3-62A IT llF IT N/A lI-3-SS CF q: NlA N/A 1I-3-208A. B CF OF NlA OF 1I-3-20ac. D CF q: NIA N/A

( d. e. f. g. OPL171.201 Revision 7 Page 33 of 117 INSTRUCTOR NOTES It is important to note that the information SER 03-05 presented in Caution #1 is not just a simple accommodation for inaccuracies in RPV water level indication which occur when plant conditions are different from those for which the instruments are calibrated. Rather, the caution defines conditions under which the displayed value and the indicated trend of RPV water level cannot be relied upon. Part B of Caution #1 identifies the limiting SER 03-05 conditions beyond which water in instrument legs may boil. Water in the RPV water level instrument legs is maintained in a liquid state by cooling action of the surrounding atmosphere and pressure in the reactor vessel. Water in the instrument legs will boil, however, if its temperature exceeds saturation temperature for the existing RPV pressure. Boiling is a concern in both horizontal and SER 03-05 vertical reference and variable instrument leg runs. Boil-off from reference leg water inventory reduces the reference head of water, decreases dP sensed by the instrument, and results in an erroneously high indicated RPV water level. Boiling in the instrument's variable leg exerts increased pressure on the variable leg side of the dP cell. This effect results in a lower sensed dP and an erroneously high indicated RPV water level. Part B of Caution #1 references the RPV SER 03-05 Saturation Temperature Curve (Curve 8) The RPV Saturation Temperature Curve is generic, based simply on the properties of water. The axis for RPV pressure is plotted from atmospheric pressure to the pressure setpoint of the lowest lifting MSRV. Note that the temperature axis of the RPV Saturation Temperature Curve is not simply drywell temperature. Depending upon the relative location of instrument reference legs and variable legs, indications from monitors near instrument runs must be considered.

( h. i. j. k. Because BFN does not have the capability of directly reading temperature indications near instrument runs located in secondary containment, the RPV Saturation Temperature Curve (Curve 8) is supplemented with Table 6, Secondary Containment Instrument Runs. Table 6 identifies the temperature elements and general locations for the instrument runs to each RPV water level instrument. Caution 1 part B says instruments "may be unreliable" if Curve 8 is exceeded. This means instruments may continue to be used until and (Gnless errat@indication is observed since momentary excursions (expected in some post LOCA situations) into curve 8 unsafe region will not result in boiling. If, however, indications of boiling are observed then that instrument is unusable until the instrument lines can be cooled and refilled. Part A of Caution #1 allows the operator to determine if each indicated RPV water level range is reliable by being above the Minimum Indicated Level for each of a series of instrument run temperature ranges. Engineering calculations have determined that when indicated RPV water level is above the Minimum Indicated Level, the operator is assured that actual RPV water level is above the instrument variable leg tap, and trends are valid. The Minimum Indicated Level is defined to be the highest RPV water level instrument indication which results from off-calibration instrument run temperature conditions when RPV water level is actually at the elevation of the instrument variable leg tap. Separate levels are provided for each RPV water level instrument. OPL171.201 Revision 7 Page 34 of 117 INSTRUCTOR NOTES The instrument will indicate high by the amount of this offset throughout its range. SER 03-05

) E MINATION REFERENCE (JPROVIDED TO CANDIDATE ( )

( CAUTIONS CAUTION #1

AN RPVWATERLVLINSTR ME

MAVeeUSB> TODETERMI EORTRENOL'V1..0 Y~ITREAOSABOVE THE MINIM M INDICATID l Vl.ASSOCIATED WITH E IGHEST MAX OWOR se R MP*

IF OW TEMPS.OR se AREA EMPS(TABlE 6). AS APPUCA8LE, ARE OUTStOE HESAFEREGIO of CURVE 8.

E ASSOCIA: 0 I liME MAYBE NRElJABLE0 E 0 BOIU G I ER MINIMUM MAX OW RUN TEMP MAXSC INSTRUMENT RANGE INDICATED (FROM XR-64-50 RUN TEMP I.Vl OR TI.-64-52AB) (FROM TABLE 6) ON SCALE N/A BELOW 150 -145 N/A 151 TO 200 LJ.3.58A,B EMERGENCY -140 N/A 201 TO 250 -155 TO +60 -130 N/A 251 TO~O

120

NlA 301 TO 350 LI-3-53 ON SCAlE N/A BaOW150 U-3-60 +5 NlA 151 TO 200 l1-3-206 NORMAL +15 N/A 201 TO 250 OTO+60 LJ.3.253 +20 N/A 251 TO~O lJ.3.208A. B,C. 0 +30 N/A 301 TO 350 LJ.3.52 POST U-3-62A ACCIDENT ON SCALE N/A NlA -268 TO +32 +10 BELOW1CO NlA +15 100 TO 150 WA SHUTDOWN +3:1 151 TO 200 WA U-3-55 FLOOOUP +30 201 TO 250 WA 010+400 +40 251 TO 300 WA "'50 301 TO 350 WA +65 351 TO 400 WA

-150" TAF -162" -175" ....J W> W -200" ....J 0 W ~o -225 0 Z -250" -268" 3-LI-3-52 &62 CORRECTION CURVES - -162"=TAF (RED LINE ) -185"=MSCRWL (GREEN LINE) -200"=MZIRWL (BLUE LINE) -215"=TWO-THIRDS CORE HEIGHT (BLACK LINE) "to. i'o. "'" " ""'t-o I'to. "" """" '"' 1-0 "" t-o -' " "" ~ I' "" I' "'"" """""" -'I' "'""" -' ...... -' ...... o 100 200 300 400 500 600 700 800 900 1000 1100 REACTOR PRESSURE (PSIG) ACTUAL LEVEL -162 11 -185" -200 11 -215 11 PIP-95-64 REV. 12

14. SRO 205000A2.06 OOllC/A/f2Gl/l-AOI-74-1//205000A2.06//SRO ONLY/12/18/2007 RMS Given the following plant conditions:

Unit-1 is in Mode 4 at 195°F following a shutdown 18 hours ear1ierfor refueling.

RHR Loop II was in Shutdown Cooling when RPV level inadvertently lowered to < 0 inches.

RPV level has been recovered to +50 inches but PCIS Group 2 logic has failed to reset. Which ONE of the following describes the appropriate course of action? A!I Obtain Shift Manager's permission to defeat PCIS Group 2 logic per 1-AOI-74-1, Loss of Shutdown Cooling, and place RHR Loop II back in Shutdown Cooling. B. Defeat PCIS Group 2 logic per 1-AOI-74-1, Loss of Shutdown Cooling, and place RHR Loop II back in Shutdown Cooling. Shift Manager's permission is not required while executing actions per an approved procedure. C. Defeating PCIS Group 2 logic is not authorized following a valid isolation signal. Make preperations to enter Mode 3 while repairs are made to the PCIS logic. D. Defeating PCIS Group 2 logic is not authorized following a valid isolation signal. Utilize alternate methods of decay heat removal in accordance with 1-AOI-74-1, Loss of Shutdown Cooling. KIA Statement: 205000 Shutdown Cooling A2.06 - Ability to (a) predict the impacts of the following on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions SDC/RHR pump trips KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the appropriate procedure to mitigate and/or recover Shutdown Cooling following an event resulting in a SDC/RHR pump trip. References: 1-AOI-74-1 Leyel of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. 0610 NRC SRO Exam

REFERENCE PROVIDED: None Plausibility Analysis: ( In order to answer this question correctly, the candidate must determine the following: 1. The loss of SOC was caused by a Group 2 PCIS isolation at +2 inches due to a valid signal. 2. The valid signal SHOULD have reset after RPV level was recovered above +2 inches, therefore the PCIS isolation signal has become "false or spurious" as defined by 1-AOI-74-1. 3. Shift Manager's permission is required to defeat PCIS logic in accordance with 1-AOI-74-1. A is correct. B is incorrect. Shift Manager's permission is required to defeat PCIS logic in accordance with 1-AOI-74-1. This is plausible because the remainder of this choice is correct and other procedures typically only require Shift Manager's permission to "deviate" from an approved procedure. C is incorrect. The valid signal SHOULD have reset after RPV level was recovered above +2 inches, therefore the PCIS isolation signal has become "false or spurious" as defined by 1-AOI-74-1, and can be defeated. This is plausible because 1-AOI-74-1 makes reference to conditions where MODE 3 may be entered, and provides appropriate guidance. D is incorrect. The valid signal SHOULD have reset after RPV level was recovered above +2 inches, therefore the PCIS isolation signal has become "false or spurious" as defined by 1-AOI-74-1, and can be defeated. This is plausible because 1-AOI-74-1 makes reference to conditions where alternate methods of decay heat removal may be used, and provides appropriate guidance.

BFN Loss of Shutdown Cooling 1-AOI-74-1 Unit 1 Rev. 0000 Page 11 of 28 ( 4.2 Subsequent Actions (continued) [11] IF the loss of Shutdown Cooling is due to Group 2 PCIS AND the isolation signal fails to reset, or will NOT remain reset due to invalid and/or sporadic signals, THEN (Otherwise N/A) PERFORM the following: [11.1] VERIFY primary containment integrity is NOT required. 0 CAunON Defeating of PCIS Group 2 isolation logic for Shutdown Cooling is ONLY to be used for an ~ isolation due to a verified false or spurious signal OR an instrument failure. Allowance for , defeating these lntenocks is NOT permitted for isolations due to an actual low water level condition. [11.2] OBTAIN Shift Manager permission to defeat the Group 2 PCIS isolation logic forShutdown Cooling. 0 [11.3] DEFEAT RHR Group 2 PCIS isolation logic in accordance with Attachment 1. 0 CAunON Group 2 PCIS isolation logic is required to be operable any time primary containment integrity is required in order to meet requirements of Technical Specifications 3.3.6.1. [11.4] IF conditions change such that primary containment integrity must be reestablished, THEN IMMEDIATELY RESTORE Group 2 PCIS Isolation Logic in accordance with Attachment 1. 0

BFN Loss of Shutdown Cooling 1-AOI-74-1 Unit 1 Rev. 0000 Page 16 of 28 ( 4.2 Subsequent Actions (continued) [20] IF Step 4.2[19] will NOT maintain the Reactor depressurized, THEN (Otherwise N/A) PERFORM the following: [20.1] VERIFY Reactor vessel makeup from condensate system is available in accordance with 1-01-2. 0 [20.2] VERIFY Main Condenser vacuum is available in accordance with 1-01-66. 0 [20.3] OPEN both Inboard and Outboard MSIVs on at least one main steam line in accordance with 1-01-1. 0 [20.4] OPEN Bypass valves as necessary to maintain the Reactor depressurized and moderator temperature less than 212°F. 0 [21] IF the Cold Shutdown Condition (Mode 4 or Mode 5) cannot be maintained, THEN (Otherwise N/A) PERFORM the following: [21.1] VERIFY the Main Condenser vacuum is available in accordance with 1-01-66. [21.2] VERIFY CLOSED RPV HEAD VENT INBD(OUTBD) VALVE, 1-FCV-3-98(99). [21.3] PERFORM the required actions for systems and components required with the reactor in other than a CorcfShufCfown Condition (Mode 4 or Mode 5) by the following:

Technical Specifications

1-GOI-100-1A o o o c

BFN Loss of Shutdown Cooling 1-AOI-74-1 Unit 1 Rev. 0000 Page 60f28 4.0 OPERATOR ACTIONS 4.1 Immediate Actions None 4.2 Subsequent Actions CAUTIONS 1) Reactor vessel stratification may occur until Shutdown Cooling is restored or a Reactor Recirculation Pump is placed in service. 2) Loss of Shutdown Cooling during the first 24 hours is most critical due to massive decay heat and limitations on the RHRSW piping. If Shutdown Cooling is lost during the first 24 hours post reactor shutdown. priorities shall be placed on the recovery of shutdown cooling in an expeditious manner [BFN PER 02-{103140-000). NOTE The following systems. if available. may be used as alternate methods of decay heat removal. REFER TO the applicable Tec Spec Bases B 3.4.7. B 3.4.8. B 3.9.7. B 3.9.8 ADHR System- (0-01-72) Fuel Pool Cooling System- (1-01-78) RWCU System- (1-01-69) Ambient losses with natural or forced circulation [1] IF any EOI entry condition is met. THEN ENTER the appropriate EOI(s). (Otherwise N/A) 0 [2] NOTIFY the Shift Manager. 0 [3] IF Refueling is in progress. THEN NOTIFY the Refueling Floor SRO. (Otherwise N/A) 0 [4] REVIEW EPIP-1. Emergency Plan Classification Logic. for entry conditions. (Otherwise N/A) 0

BFN Loss of Shutdown Cooling 1*AOI*74*1 Unit 1 Rev. 0000 Page 10 of 28 4.2 Subsequent Actions (continued) [8.3] DETERMINE the reactor coolant temperature or use the last valid reactor coolant temperature available. [8.4] IF the Reactor Vessel head is removed AND the cavity is flooded with the fuel pool gates installed, THEN (Otherwise N/A) o ESTIMATE the time for reactor coolant temperature to reach 125°F and 150°F using a plot of the actual heatup rate or Illustration 1: 0 [8.5] ESTIMATE the time for reactor coolant temperature to reach 212°F, using data obtained in Steps 4.2[8.1] through 4.2[8.3]. [9] IF the loss of Shutdown Cooling is due to inadequate RHRSW flow, THEN (Otherwise N/A) o START the standby RHRSW pump for the appropriate header in accordance with 0-01-23. 0 [10] IF the loss of Shutdown Cooling is due to Group 2 PCIS isolation, THEN (Otherwise N/A) PERFORM the following when conditions permit resetting Group 2 PCIS isolation: [10.1] RESET Group 2 isolation by momentarily placing PCIS OIV I RESET, 1-HS-64-16A-S32, and PCIS OIV II RESET, 1-HS-64-16A-S33, in reset. 0 [10.2] MOMENTARILY DEPRESS RHR SYS I SO CLG INBO INJECT ISOL RESET, 1-XS-74-126 and RHR SYS II SO CLG INBO INJ ISOL RESET, 1-XS-74-132. [10.2.1] VERIFY 1-IL-74-126 and 1-IL-74-132 extinguished. 0

BFN Loss of Shutdown Cooling 1-AOI-74-1 Unit 1 Rev. 0000 Page 11 of 28 4.2 Subsequent Actions (continued) [11] IF the loss of Shutdown Cooling is due to Group 2 PCIS AND the isolation signal fails to reset, or will NOT remain reset due to invalid and/or sporadic signals, THEN (Otherwise N/A) PERFORM the following: [11.1] VERIFY primary containment integrity is NOT required. 0 CAUTION Defeating of PCIS Group 2 isolation logic for Shutdown Cooling is ONLY to be used for an isolation due to a verified false or spurious signal OR an instrument failure. Allowance for defeating these interlocks is NOT permitted for isolations due to an actual low water level condition. [11.2] OBTAIN Shift Manager permission to defeat the Group 2 PCIS isolation logic for Shutdown Cooling. 0 [11.3] DEFEAT RHR Group 2 PCIS isolation logic in accordance with Attachment 1. 0 CAUTION Group 2 PCIS isolation logic is required to be operable any time primary containment integrity is required in order to meet requirements of Technical Specifications 3.3.6.1. [11.4] IF conditions change such that primary containment integrity must be reestablished, THEN IMMEDIATELY RESTORE Group 2 PCIS Isolation Logic in accordance with Attachment 1. 0

BFN Loss of Shutdown Cooling 1-AOI-74-1 Unit 1 Rev. 0000 Page 12 of 28 4.2 Subsequent Actions (continued) [12] IF the Group 2 PCIS Isolation has been reset, THEN (otherwise N/A) RETURN the affected loop of RHR to Shutdown Cooling as follows: [12.1] CLOSE RHR SYS 1(11) LPCI OUTBD INJECT VALVE, 1-FCV-74-52(66). 0 [12.2] OPEN RHR SYS 1(11) LPCI INBD INJECT VALVE, 1-FCV-74-53(67). 0 [12.3] VERIFY RHR SYSTEM 1(11) MIN FLOW INHIBIT Switch 1-HS-74-148(149) in INHIBIT. 0 [12.4] VERIFY CLOSED RHR SYSTEM(SYS) 1(11) MIN FLOW VALVE, 1-FCV-74-7(30). 0 [12.5] VERIFY CLOSED RHR PUMP 1A(1B) and 1C(1D) SUPPR POOL SUCT VLVs, 1-FCV-74-1(24) and 1-FCV-74-12(35). 0 [12.6] VERIFY OPEN RHR PUMP 1A(1B) and 1C(1D) SD COOLING SUCT VLVs, 1-FCV-74-2(25) and 1-FCV-74-13(36). 0 [12.7] OPEN RHR SHUTDOWN COOLING SUCT OUTBD and INBD ISOL VLVs, 1-FCV-74-47 and 1-FCV-74-48. 0 [12.8] RESTART RHR PUMP 1A(1C)(1B)(1D) using 1-HS-74-5A(16A)(28A)(39A). 0

Given the following plant conditions: 15. SRO 212000A2.12 OOIlC/A!f2Gl/I/212000A2.12/1SRO ONLYIl21l8/2007 RMS (

Unit-2 is at 60% rated power. An Instrument Mechanics report that PIS 1-91A and 1-81A were discovered to have failed upscale. ~

Further investigation determines that these two pressure switches feed the Turbine Control Valve/Stop Valve Closure Trip Bypass logic for RPS. Which ONE of the following describes the the status of the RPS logic with respect to TCVlSV closure scram capability and the required actions per Technical SpecificationsyTf'l../AIL 7 ~ I REFERENCE PROVIDED (l1S.8 ~ A TCVlSV Closure Scram capability is maintained. Restore PIS 1-91A and 1-81A to operable status in 1 hour or initiate insertion of OPERABLE rods immediately. B~ TCVlSV Closure Scram capability is maintained. Initiate an INFORMATION ONLY LCO. If power is reduced below 30% rated power, place RPS -B- in trip within 1 hour. I C. TCVlSV Closure Scram capability is NOT maintained. Reduce power below 30% rated power within 4 hours. D. TCVlSV Closure Scram capability is NOT maintained. Remove power from PIS 1-91A and 1-81A within 1 hour. KIA Statement: 212000 RPS A2.12 - Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Main turbine stop control valve closure KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the appropriate status and required actions for a Main turbine stop control valve closure logic failure. References: Level of Knowledge Justification: This question is rated as CIA due to the requirement to assemble, sort, and integrate the parts of the question to predict an outcome. This requires mentally using this knowledge and its meaning to predict the correct outcome. SRO Level Justification: This question satisfies the requirements of 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases. 0610 NRC SRO Exam }}

ML081370271

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