IR 05000259/2020008
| ML091130435 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 02/25/2008 |
| From: | NRC/RGN-II |
| To: | Tennessee Valley Authority |
| References | |
| 50-259/08-301, 50-260/08-301, 50-296/08-301 50-259/08-301, 50-260/08-301, 50-296/08-301 | |
| Download: ML091130435 (386) | |
Text
{{#Wiki_filter:1. RO 203000A1.0 1 00 lIelAlT2G 1IRHRlDWSP/1I203000A1.0 1I4.2/4.3/ROIMODIFIED 11117/07 Given the following conditions:
* Unit 2 has experienced a Loss of Coolant Accident (LOCA).
- Drywell sprays are required in accordance with 2-EOI-2 flowchar Which ONE of the following plant conditions must exist prior to opening BOTH the Residual Heat Removal (RHR) SYS I Inboard AND Outboard Drywell Spray Valves? RPV level must be greater than (-)155 inches (Emergency Range) with ONLY the the CO NT SPRAY VLV SEL SWITCH in SELEC RPV level must be greater than (-)162 inches (Post Accident Range) with ONLY the CONT SPRAY VL V SEL SWITCH in SELEC C." RPV level is greater than (-)183 inches (Post Accident Range) with ONLY the CO NT SPRAY VL V SEL SWITCH in SELEC RPV level is less than (-)200 inches (Post Accident Range) with ONLY the 2/3 CORE HEIGHT KEYLOCK BYPASS SWITCH in BYPAS KIA Statement:
203000 RHRlLPCI: Injection Mode A 1.01 - Ability to predict and/or monitor changes in parameters associated with operating the RHRlLPCI: INJECTION MODE (PLANT SPECIFIC) controls including: Reactor water level KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific values of reactor water level to determine the conditions which allow diverting RHR from a LPCI Injection lineup to containment contro 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 outcom References: 2-01-74, OPL 171.044, TP-33 0610 NRC RO Exam modified from OPL 171.044 #22 REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: Page: lof3 2/19/2008
1. Drywell sprays being required infers that DW pressure is >2.45 psi . Based on the RPV level conditions given in the available answers, determine whether a CAS signal has been generated due to the LOC . Which switch(s) must be manipulated to override a CAS signal with the existing condition NOTE: All of these answers are plausible based on minimal procedural guidance given in EOI Appendix 17B. Experience has shown that both switches are manipulated by novice operators regardless of conditions to facilitate Drywell sprays as required. This is not a procedure violation, but demonstrates a lack of specific knowledge of required condition C - correct. OPL 171.044, TP 33 shows that under the conditions only the CONT SPRAY VL V SEL SWITCH in SELECT is required to allow spraying containmen Contact K 14A1B is closed if level is above 2/3's core height (-183") as referenced on the Post Accident range (-268" to +60"). Contact K61A1B is closed due to an accident signal (level <-122" on the Wide range level instruments). Therefore to complete the logic only the SELECT switch is required to be closed which energizes relays K58A1B (Seal In) and K50A/B which allows spraying containmen A - incorrect. With RPV level> -155 inches (-155" bottom of scale on the emergency range is operationally considered top of active fuel) adequate core cooling is assured by submergence and therefore spraying containment is allowed. If candidate determines that top of active fuel verses 2/3 core height is necessary, it becomes ( plausible. Also level must be above -183" not -155" B - incorrect. With RPV level> -162 inches (top of active fuel) adequate core cooling is assured. If candidate determines that top of active fuel versus 2/3's core height is necessary, it becomes plausible. Also level must be above -183" not -162" o - incorrect. With RPV level < -200 inches, you are below 2/3's core height which requires both the CO NT SPRAY VL V SEL SWITCH in SELECT and 2/3 CORE HEIGHT KEYLOCK BYPASS switch in BYPAS Amplification: In regard to specific comments/questions,concerning valve numbers versus noun names at BFN, this poses a problem for systems with multiple loops such as RHR, Core Spray and Recirc. TVA does NOT use conventional GE Master Parts List (MPL) numbers like most BWR plants. The same valve NAME on different loops will have a different NUMBER. In this question it doesn't matter which "loop" we are referencing since the interlock in question is the same. Therefore, using the noun name is appropriate without the number. In other cases where there is only one valve by that NAME, using the number AND the name will reduce confusio Regarding RPV level instrumentation, all GE BWRs have multiple ranges of instrumentation to cover the large height of the reactor vessel and maintain accuracy under various operating conditions within the vessel. This design feature, unique to BWRs, creates a significant challenge to our operators. All of the various ranges of RPV level instrumentation have control functions that input to various systems such as RPS, PCIS, LOCA logic and ECCS initiation logic. As such, providing the candidate with level indications, with or without the corresponding name of the range in question, Page: 2 of 3 2/1912008
is all part of determining whether that candidate possesses the required knowledge of RPV level instrumentation and its effect on plant operatio Browns Ferry has the following level ranges: Narrow range (Normal Control range): 0" to +60" Wide range (Emergency Systems range): -155" to +60" Shudown Vessel Floodup (Floodup range): 0" to +400" Post Accident Flood range (Post Accident range): -268" to +32" The 2/3 core height logic is developed using the Post Accident Flood range instrumentation, whereas the accident signal logic is developed using the Wide range instrumentation.
( Page: 3 of3 2/1912008
QUESTIONS REPORT for 0610 NRC RO Exam I* Select __ XS-74-121 1I Select, N AS (129) K58A (5S8) Closed if
250 VDC K14 A Water Level Manual (148) 2/3 Core ________Q~!rj~! _____________ Height XS-74-122 (130) K61A (61 B) K58A K50A (588) (SOB) Aux Relay Select Loglcs to Valves 74-59 (73) and 74-57 (71) Select XS-74-121 (129) I Select / N AS K69 Seal in K60A K60B 250 VDC Closed if Dry 'W' e 1T K1 00 A Pressure K1 OOB J; 1 1 .96 psig Closed if t - - - - - - - - - 1
~
Ma'riual
'Water LE'YE'l IK14A (148) - -Ov;rid-; - - - - - --
2/3 Core Height 74-122 K61A (618) (74-130) K69A (698) K59A. (598) Aux RE'lay SE'lect Logics for Containment/Torus Spray Valves 0610 NRC RO Exam modified from OPL 171.044 #22 Thursday, February 07, 2008 7:55:22 AM 2
BFN Residual Heat Removal System 2-01-74 Unit 2 Rev. 0133 Pa_ge 23 of 367 INTERLOCKS (continued) The RHR spray/cooling valves, 2-FCV-74-57(71), receive an auto closure signal in the presence of a LPCI initiation signal and they are interlocked to prevent opening if the in-line torus spray valve, 2-FCV-74-58(72), is not fully closed. The in-line valve interlock can be by-passed if the following conditions exis (1) Reactor level is >2/3 core height and a LPCI initiation signal is present and the select reset switch is in the SELECT positio The requirements for >2/3 core height and a LPCI initiation signal may be by-passed using the keylock bypass switch, 2-XS-74-122/3 . If primary containment cooling is desired with reactor level at <2/3 core height, the keylock bypass switch is required to be placed in BYPASS before the select reset switch is placed in SELECT to ensure relay logic is made u . The RHR torus spray valves, 2-FCV-74-58(72), have the same in-line valve interlocks as those outlined in Step 3.5A.8 for the torus spray/cooling valves. Additionally these valves have an interlock preventing opening unless drywell pressure is 21.96 psig which cannot be bypasse . The RHR torus cooling/test valves, 2-FCV-74-59(73), receive an auto closure signal in the presence of a LPCI initiation signal. Auto closure may be bypassed by the same conditions/actions outlined in Step 3.5 . The RHR containment spray valves, 2-FCV-74-60(74) and 61 (75), have in-line valve interlocks similar to these described in Step 3.5 through 3.5A.1 0 for the RHR torus spray valves 2-FCV-74-57(58) and 71 (72).
12. If 2-FCV-74-59(73) LOCA CLOSURE TIME light (2-IL-74-59Y Loop I; 2-IL-74-73YLoop II) on Panel 2-9-3 is extinguished due to its associated valve being opened, that Loop is inoperable for LPC . If2-HS-74-148(149) RHR SYSTEM I (II) MIN FLOW INHIBIT switch is in the INHIBIT position, the pumps on that loop do not have automatic minimum flow protectio EOI APPENDIX-17B Rev. 10 Page 2 of 13 6. INITIATE Drywell Sprays as follows: VERIFY at least one RHRSW pump supplying each EECW heade IF ..... EITHER of the following exists:
* LPCI Initiation signal is NOT present, OR * Directed by SRO, THEN ... PLACE keylock switch 2-XS-74-122(130), RHR SYS 1(11) LPCI 2/3 CORE HEIGHT OVRD, in MANUAL OVERRID MOMENTARILY PLACE 2-XS-74-121(129) , RHR SYS 1(11)
CTMT SPRAY/CLG VLV SELECT, switch in SELEC IF ..... 2-FCV-74-53(67), RHR SYS 1(11) LPCI INBD INJECT VALVE, is OPEN, THEN ... VERIFY CLOSED 2-FCV-74-52(66) , RHR SYS 1(11) LPCI OUTBD INJECT VALV VERIFY OPERATING the desired System 1(11) RHR pump(s) for Drywell Spra f. OPEN the following valves:
* 2-FCV-74-60(74) , RHR SYS 1(11) DW SPRAY OUTBD VLV * 2-FCV-74-61 (75), RHR SYS I (II) DW SPRAY INBD VL VERIFY CLOSED 2-FCV-74-7(30) , RHR SYSTEM 1(11) MIN FLOW VALV IF ..... Additional Drywell Spray flow is necessary, THEN ... PLACE the second System 1(11) RHR Pump in servic MONITOR RHR Pump NPSH using Attachment j. VERIFY RHRSW pump supplying desired RHR Heat Exchanger(s) .
2. RO 205000K4.02 001lMEM/T2/Gl/RHRJ/205000K4.02/3.7/3.8/RO/BANK 11127/07 RMS Given the following conditions on Unit 2:
* Reactor level (+ )20 inches * Reactor pressure 90 psig * Drywell pressure 1.7 psig Which ONE of the following describes which modes of Residual Heat Removal (RHR)
are available for use (consider interlocks only)? Low Pressure Coolant Injection (LPCI), Drywell Sprays, and Shutdown Coolin Suppression Pool Cooling, Suppression Pool Sprays, and Shutdown Coolin C." Suppression Pool Cooling, Low Pressure Coolant Injection, and Shutdown Coolin Low Pressure Coolant Injection (LPCI), Drywell Sprays, and Supplemental Fuel Pool Coolin KIA Statement: 205000 Shutdown Cooling K4.02 - Knowledge of SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) design feature(s) and/or interlocks which provide for the following: High pressure isolation: Plant-Specific KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine which interlocks apply to those conditions including the High Pressure isolation of RHR SDC mod References: OPL 171.044 rev 15, pg 31, C -Instrumentation and Interlocks, 2-01-74-rev 133, step 8.8[11] Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam Page: 1of2 2/19/2008
REFERENCE PROVIDED: None ( Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Orywell pressure <1.96 psig prohibits use of containment sprays (OW and SP).
2. RPV pressure <450 psig aHows the use of LPCllnjectio . RPV pressure of 90 psig allows the use of Shutdown Coolin NOTE: All given answers are plausible since they all contain at least one acceptable lineup with the given condition C - correct. The lack of an accident signal (an accident signal is either +2.45 psig High Orywell pressure and <450 psig reactor pressure or -122" low reactor water level) allows Suppression Pool Cooling. Reactor pressure <450 psig allows Low Pressure Coolant Injection. Reactor pressure at 90 psig is below the high pressure isolation setpoint of 100 psig for Shutdown Cooing and Orywell pressure is below the PCIS group II isolation setpoint of +2.45 psig; therefore, Shutdown Cooling is availabl A - incorrect. LPCI and Shutdown Cooling would be allowed by interlocks. Orywell Sprays will not function <1.96 psig in the Orywel B - incorrect. Suppression Pool and Shutdown Cooling would be allowed by interlocks. Suppression Pool Sprays will not function <1.96 psig in the Orywel o - incorrec LPCI and Supplemental Fuel Pool Cooling would be allowed by interlocks. Orywell sprays will not function, as previously state Amplification: In responding to your specific questions, concerning "Consider interlocks only", the reason for that statement is that the given pressure is procedurally too high to place RHR in Shutdown Cooling, but the interlocks will allow it. Per 01-74, Section 8.8, "VERIFY Reactor pressure is less than 55 psig, OR if entering this procedure from RC/P of 2-EOI-1, pressure is less tha~ 100 psig."
In question #1, the allowance or preclusion of Orywell Sprays are related to reactor vessel water level. In this question, it is a Orywell pressure setpoint, versus RPV water level, which is associated with answering the question correctly. If the candidate only assumes that the + 20 inches RPV water level will allow sprays, then the candidate will answer the question incorrectl Additionally, a specific instrument is not required to answer this question, since the level that is given is not above or below any initiation or trip setpoin In order to properly comply with the KIA statement, the candidate's knowledge of the high pressure INTERLOCK associated with RHR Shutdown Cooling must be evaluated, not the procedural limitations. A pressure above the procedure limit but BELOW the interlock for was given for that purpos Page: 20f2 2/19/2008
BFN Residual Heat Removal System 2-01-74 Unit 2 Rev. 0133 Page 107 of 367 Initiation/Operation of Loop 1(11) Shutdown Cooling (continued)
[10.2] IF RECIRC PUMP 2B(2A) SUCTION VALVE, 2-FCV-68-77(1) is to be closed, THEN PERFORM the following: [10.2.1] VERIFY TRIPPED, RECIRC DRIVE 2B(2A)
NORMAL FEEDER, 2-HS-57-14(17). o
[10.2.2] VERIFY TRIPPED, RECIRC DRIVE 2B(2A)
ALTERNATE FEEDER, 2-HS-57-12(15). o
[10.2.3] CLOSE tripped recirc pump suction valve using RECIRC PUMP 2B(2A) SUCTION VALVE, 2-HS-68-77(1 ). o [11] VERIFY Reactor pressure is less than 55 psig, OR if entering this procedure from RC/P of 2-EOI-1, pressure is less than 100 psi [12] DIRECT Instrument Mechanics to enable RHR SD CLG FLOW LOW annunciator, 2-XA-55-3D, Window 11 and VERIFY setpoint of 3700 gpm by programming recorder 2-FR-74-64, RHR SYS 1/11 FLOW, for RHR Loop to be placed in Shutdown Coolin [13] NOTIFY Chemistry that RHRSW is to be placed in service and Shutdown Cooling is to be starte NOTES 1) For closed loop vents, venting is required for 1 minut ) Step 8.8[14] may be N/A'd, when the RHR Loop has been vented within 24 hour [14] OPEN the following RHR Loop 1(11) vent valves until a solid stream of water is observed, THEN CLOSE: Head (Containment) Spray Line through RHR SYS I HEAD SPRAY HI POINT (RHR SYS DW SPRAY)
TELL-TALE VENT SOV, 2-SHV-074-0746(0747), AND 0 HIGH POINT TELL TALE VENT HEAD SPRAY LINE (CONTAINMENT SPRAY), 2-FSV-74-138(139). [Rx Bldg, EI 593' Fuel Pool Cooling Area (Rx Bldg, E, EI 621')] 0
OPL 171.044 Revision 15 Page 31 of 159 INSTRUCTOR NOTES c. Reactor pressure interlock 100 psig TP-1 and 2 Two pressure switches monitor reactor system PS-68.;93 pressure at the "A" Reactor Recirculation Pump PS-68-94 Suction and prevent opening (automatic-closure) the 74-48 & 47 isolation valves unless sensed pressure is <100 psig (greater than 100 psig is beyond RHR piping design). These press switches do NOT provide the alarm function Orywell pressure 2.45 psig (1) Pressure switches initiate LPCI mode of RHR in conjunction with < 450 psig reactor pressure, interlock Containment Cooling/spray valves closed, initiate PCIS Group 2 isolation and initiate a reactor scra (2) The OW pressure switches operate relays in the RHR Logic system (Le. does not work thru Core Spray) Reactor vessel water level -183 inches indicated on Post Accident Flooding Range LIS used to provide a level permissive signal to LlS-3-52 & 3-62A Containment Cooling valves based on level inside the shroud. Permits containment spray only if level is >2/3 core height with LOCA signal presen * A keylock bypass switch can bypass the 2/3 core height interlock. Orywell pressure 1.96 psig Pressure switches used to provide pressure Monitoring plant permissive signal to containment spray valve systems is critical Permits opening of containment spray only if during a transient pressure is significant in containment after acciden when conditions change rapidl * OWP decreasing to less than 1.96 psig will SER 03-05 automatically close the spray valves if an accident signal is present. Reactor vessel low water level +2 inches (Lvi 3) Level switches used to initiate: PCIS Group 2 isolation of Shutdown Cooling, CS System, RHR SOC isolation valves, RHR inbd isolation valves, RWCU isolation and a reactor scra OPL 171.044 Revision 15 Page 40 of 159 INSTRUCTOR NOTES (3) . Can be opened when Rx press is ~ 450 psi NOTE: Effect of Logic failure and valve operation (4) . Automatically opens on LPCI initiation signal when reactor pressure is < 450 psi (5) LPCI Injection Valve Open Signal Bypass NEW! Switch (Keylock switch on 9-3) can be utilized 10A-S155A(B) to bypass the open signal during execution of Unit 1 and 2 EOI's. Allows operator to manually (pnl 9-3) ONLY at this time close the injection valve Indicating light (6) The normally open outboard injection valves informs operators (1-FCV-74-52,74-66;2-FCV-74-52, 74-66; 3-74-when open signal 66) have added circuitry so that a fire cannot logic is bypasse energize the closing coil and shut the valve (any close signal with the Control Room handswitch in NORMAL. Shorts out the closing coil and blows the control power fuses).
Modifications also disabled the local control
"Close" pushbutton on 1-FCV-74-52/66, 2-FCV-74-52/66 and 3-FCV-74-6 (7) Control Circuit
, , 1-74-52 1-74-66 2-74-52 2-74-66 3-74-52 3-74-66 J Operate Interlock 74-53 74-67 74-53 74-67 74-53 74-67 I
--j i Outgoing interlock 74-53 74-67 74-53 74-67 74-53 74-67 I Normal/Emerg Sw X X X I Local Controls X , X X X X X I Controls at Bkr X X X Panel 9-3 controls X X X X X
;
i X ~ I I Local Indication X X X X X i X I I Igts i I
~-!
I
,
!Lights on Bkr r;---:-" I I X ! X X , ~----~-~, i I L1~ghts on Pnl 9-3 X X X X X X LPCI inboard injection valves (74-53; 74-67) Obj. V. (1) Normally closed - non-throttling TP-29, 30, and
OPL 171.044 Revision 15 Page 41 of 159 INSTRUCTOR NOTES (2) Interlock prevents normal opening unless in-line valve (74-52; 74-66) is fully closed with reactor pressure> 450 psig Operation of the valve at the breaker using the 1-74-53 only controls there will bypass the in-line and 450 1-74~67 only psig interlock; prevents automatic opening and 2-74-53 only closure due to logic; and prevent any operation 3-74-53 only except from breaker (3) Can be opened when Rx pressure is < 450 NOTE: Effect of psig Logic failure and valve operation (4) Automatically opens when Rx pressure < 450 NEW! psig with an LPCI initiation signal present and The Redundant is interlocked open until LPCI initiation signal is logic has been cleared and rese remove (5) Only Respective Divisional LPCI Initiation logic will close the valv (6) Automatically close (both valves) if: ( FCV 74-47 and 48 (SID Cooling supply valves) open and a Group 2 isolation signal occurs Automatic closure signal seals in (light Sys I-XS-74-126 indication). Can be reset (FCV 74-53/67 Sys II-XS-74~132 Shutdown Cooling isolation reset push buttons) when any of the conditions above are cleare Note that this closure signal will prevent opening if an LPCI signal is receive (7) The normally closed inboard injection valves (2/3-FCV-74-53 and 74-67) have a new App 'R' Emergency Open Switch on the power supply board to bypass all interlocks and other circuitry (except the fully open limit switch) to open the valv OPL 171.044 Revision 15 Page 43 of 159 INSTRUCTOR NOTES (7) Separate bypass switch allows bypassing interlock from Valves 74-2/13 (74-25/36)
(8) Control Circuit 1-74-57 1-74-71 2-74-57 2-74-71 3-74-57 3-74-71 Operate Interlock 74-58, 74-72, 74-58, 74-72, 74-58, 74-72, 74-74-2/13 74-25/36 74-2/13 74-25/36 74-2/13 25/36 Outgoing interlock 74-2/13 74-25/36 74-2/13 74-25/36 74-2/13 74-25/36 Normal/Emerg Sw X X X Local Controls X X X X X X Controls at Bkr X X X Panel 9-3 controls X X X X X X Local Indication X X X X X X Igts Lights on Bkr X X X Lights on Pnl 9-3 X X X X X X Bypass Switch X X X X X X m. RHR Suppression Pool spray valves (74-58; 74-72)
Obj. V.C.5.
( (1) No automatic opening logic TP-33, 36 and
(2) Interlock prevents normal opening if in~line valve not full closed (74-57; 74-71) (3) Automatically closed and interlocked closed on LPGI initiation signa (4) The in-line valve interlock and/or the LPCI closure signal can be bypassed if the following exist: (a) Reactor level .:::.-183 inches and drywell pressure.:::. 1.96 psig and LPCI initiation signal and Select-Reset switch to SELECT positio (b) Reactor level interlock and LPCI initiation signal may be bypassed by use of keylock bypass switch (XA 74-122/130)
OPL 171.044 Revision 15 Page 44 of 159 INSTRUCTOR NOTES (5) Amber light above the "SELECT" switch Obj. V.B.11 indicates: Obj. V. Switch in "Select or Normal after Select" AND DWP is .::1.96 psig AND RPV level .::-183" & have LPCI signal OR Keylock in Bypass position (a) As long as the light remains "on", the valves may be opened and a LPCI signal will not close the (b) The in-line interlock valve does not have the DWP interlock, so it is possible to open the Spray valve (58/72) first and then open the in-line valve (57/71) without DWP being .::1.96 psig (6) Drywell pressure interlock prevents drawing This interlock vacuum on containment under accident cannot be conditio bypasse (7) Control Circuit I i 1-74-58 1-74-72 I 2-74-58 I 2-74-72 I 3-74-58 3-74-72 I I I I Operate Interlock 74-57 74-71 74-57 74-71 74-57 74-71 Outgoing interlock NormallEmerg Sw Local Controls X X X X X X i Controls at Bkr I Panel 9-3 controls X X X X X I X I Local Indication Igts X X X X X X I . Lights on Bkr i L Lights on Pnl 9-3 X X X X X X
OPL 171.044 Revision 15 Page 46 of 159 INSTRUCTOR NOTES Containment Spray valves (74-60/61 ; 74-74/75)
(1 ) No automatic opening logic Obj. V. (2) IN-line valve interlock prevents normal opening TP-33, 40, 41, unless other valve fully closed 42,43 (3) Automatically closed/interlocked closed on LPCI signal (4) Automatic closure signal and/or the in-line valve interlock may be bypassed if the following exist: (a) Reactor level,::-183 inches and drywell pressure ,::1.96 psig and LPCI initiation signal present and Select=Reset switch placed to SELECT position (b) Reactor level interlock and LPCI initiation signal may be bypassed by use of keylock bypass switch (XS-74-122/130) (5) Amber light above the "SELECT" switch Obj. V. indicates:
Switch in "Select or Normal after Select" AND DWP is ,::1 .96 psig AND RPV level,::-183" and have LPCI signal OR Keylock in Bypass position (a) As long as the light remains "on", the valves may be opened and a LPCI signal will not close the (6) Drywell pressure interlock prevents drawing This interlock vacuum on containment under accident cannot by conditio bypasse (7) Emergency position at breaker bypasses both Obj. V. of the normal control circuits U2 & U3-74-60 (opening/closing/interlocks) U1-74-74
3. RO 206000K6.09 001/C/A/SYS/HPCI/41206000K6.09//RO/SRO/BANK 11/27/07 RMS Given the following plant conditions:
* Unit 2 reactor water level initially lowered to (-)69 inche * Conditions required entry into EOI-1, "RPV Control" and EOI-2, "Primary Containment Control."
- After water level recovery, the High Pressure Coolant Injection (HPCI) Pump Injection Valve (FCV-73-44) was manually closed and HPCI was placed in pressure control to remove decay hea * Subsequently, Condensate Storage Tank (CST) level drops below 6800 gallon Which ONE of the following describes the status of the HPCI system (assume no other operator actions have occurred)? HPCI would be operating in pressure control with suction from the CS HPCI would be pumping to the CST with suction from the Suppression Poo C. \I' HPCI would be operating at shutoff head with suction from the Suppression Poo The HPCI turbine would trip on overspeed due to loss of suction during the transfe KIA Statement:
206000 HPCI K6.09 - Knowledge of the effect that a loss or malfunction of the following will have on the HIGH PRESSURE COOLANT INJECTION SYSTEM: Condensate storage and transfer system: BWR-2,3,4 KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the effect of low CST level on HPCI operatio References: OPL 171.042 Rev 19 Page 36 and TP-12 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 outcom NRC Exam Page: lof2 2/19/2008
REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Recognize that the HPCI initiation signal is reset to allow HPCI to be placed in Pressure Contro . Recognize that the HPCI Pressure Control lineup if from the CST and back to the CS . Recognize that the current CST level would initiate a suction swap to the Suppression Poo . Recognize that HPCI would not receive a trip signal as the suction valves re-aligne . Recognize that the CST Test Isolation Valve will auto close on low CST leve C - correct: At -7000 gallons in the CST, HPCI auto swaps from CST suction to Suppression Pool (Torus) suction. When this occurs the CST Test Return Isolation valve receives a close signal from the Torus suction valves opening; to prevent pumping the Torus to the CST. Therefore, with the HPCI injection valve previously closed, HPCI would be operating at shutoff head without minimum flow protectio Note - the minimum flow valve would be closed due to lack of a valid initiation signa A - incorrect: This assumes the low CST level has NOT initiated a suction swap to the Torus. This is plausible since most procedures list the setpoint for the auto swap as an elevation above sea level versus a "gallons" setpoin B - incorrect: This lineup would occur if the HPCI Test Isolation Valve did NOT receive a close signal following the suction swap logic initiation. This is plausible since the ONLY auto closure interlock of the HPCI Test Isolation Valve is under this specific conditio D - incorrect: HPCI will not trip on low suction pressure under this specific conditio The Torus suction valves begin to open before the CST suction valve closes in a
"make-before-break" fashion. This is plausible since closure of the suction path to HPCI typically results in a low suction trip.
Page: 2of2 2119/2008
OPL 171.042 Revision 19 Appendix C Page 66 of 67 TP-12: HPCI Valve Logic (2 of 3)
OPL 171.042 Revision 19 Page 36 of 67 INSTRUCTOR NOTES 2. If during HPCI operation, suppression pool water level increases to 7" (5.2" on Unit 3) above zero or if CST level drops to 552'6" above sea level (7000 gallons), then HPCI pump suction valves from the suppression pool (73-26 and 73-27) open. (This will then cause the CST suction valve to close once the SP suction valves get full open).
NOTE: There are normally 300,000 gallons available in the CST for HPCI and RCIC use.
3. A flow switch tapped in parallel with the HPCI system flow controller closes the minimum flow bypass valve to suppression pool (73-30) at 1255 gpm increasing; and opens it at 900 gpm decreasing, only if an auto start signal is presen Minimum flow valve closes on a Turbine Trip signal.
4. If either of the suppression pool suction line Obj. V.8A isolation valves (73-26 or 73-27) are full open then Obj. V.CA the HPCI test line to the CST valves (73-35 and 73-36) will close.
5. If the HPCI turbine isolation valve (73-16) is fully closed, then gland seal condenser condensate pump discharge valves to clean radwaste (73-17 A and 73-178) will open if the gland seal condenser hotwell has high level.
6. If the HPCI turbine isolation valve (73-16) is fully closed, then HPCI turbine steam line drain pot discharge isolation valves to the main condenser (73-6A and 73-68) will open.
7. If 73-16 is full closed, the auxiliary oil pump will not start from the control room. When 73-16 opens 10% and the control switch is in the start position, the auxiliary oil pump will run.
8. If the HPCI turbine steam line drain pot level DCN 51221 reaches the high level setpoint, then the downstream trap bypass valve (73-5) will ope Unit difference Unit 1 73-5 has been replaced with a manual valv . RO 209001K5.04 001IMEMlT2G1IBASISI1209001K5.04//RO/SROIMODIFIED 11117/07 During EOI execution, when injection from low pressure systems is required to restore and maintain RPV level, the Core Spray System is NOT on the list of preferred ( systems for low pressure injection IF all control rods are NOT inserte Which ONE of the following describes the basis for this restriction? Cold water from Core Spray creates a rapid pressure reduction and cooldown rates CANNOT be controlle B." Core Spray injects directly onto fuel bundles inside the shroud which could damage fuel and cause a power excursio Core Spray injection creates a steam blanket at the top of the fuel bundles which inhibits heat transfer via steam flow past the fue Core Spray does NOT provide sufficient flow to maintain adequate core cooling if an ATWS power level greater than or equal to 80% occur KIA Statement: 209001 LPCS K5.04 - Knowledge of the operational implications of the following concepts as they apply to LOW PRESSURE CORE SPRAY SYSTEM: Heat removal (transfer) mechanisms KIA Justification: This question satisfies the KIA statement by requiring the candidate to recall the unique heat removal mechanisms of Core Spray and recall a condition where that mechanism can result in unfavorable consequence References: OPL 171.205 rev 8, pg 60, 11.d and EOIPM Section O-V-K Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam MODIFIED FROM OPL 171.205 #9 Page: lof2 2/19/2008
REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. The bases behind the restriction of Core Spray injection during an ATWS emergenc B - correct: The systems listed in Step CS-13 for use in controlling RPV water level comprise all those which inject outside of the core shroud. These are used, preferentially, because the flow path outside the core shroud mixes the relatively cold injected water with the warmer water in the lower plenum prior to it reaching the cor Core Spray injects outside the shroud spraying relatively cold water directly on the fuel which does not provide mixing of the boron (if injected) could lead to fuel damage or a power excursio A - incorrect: This is could cause a rapid pressure reduction but is not the overriding factor for not using Core Spray under these conditions. This is plausible since high volume Core Spray injection at close to the maximum injection pressure would cause a rapid pressure reductio C - incorrect: This phenomenom, referred to as Counter Current Flow Instability, is plausible but is only of significant concern with the core completely uncovered and is the basis for removing Spray Cooling from the EPG definition of Adequate Core Coolin D - incorrect. This is incorrect because Core Spray can provide enough flow to maintain adequate core cooling at any ATWS power level if water level is lowered lAW CS. This is plausible since Core Spray would not be able to maintain water level if level was not lowered in CS.
Page: 20f2 2119/2008
C5, LEVEUPOWER CONTROL BASES EOI PROGRAM MANUAL SECTION O-V-K I:: DISCUSSION: STEP C5-16 (Continued)
I In comparison to Minimrun Zero-Injection RPV Water Level (refer to the discussion of Step C3-3 in the C3, Steam Cooling, Bases), Minimrun Steam Cooling RPV Water Level is slightly higher than Minimum Zero-Injection RPV Water Level. This is attributed to two key factors:
, Injection of subcooled water requires that part of the energy that would be used to generate steam for cooling the uncovered portion of the core must now be expended in heating sub cooled liquid to saturation temperature (Minimum Zero-Injection RPV Water Level is calculated assuming no injection into the RPV). More steam is required to maintain clad temperature below 1500 OF as compared to the 1800 OF limit assruned for Minimum Zero-Injection RPV Water Level calculatio The injection sources listed for use in controlling RPV water level comprise all of those that inject outside of the core shroud. These are used, preferentially, because the flowpath outside the core shroud mixes the relatively cold injected water with warmer water in the lower plenum prior to reaching the core. No priority between use of each listed system is intended, therefore the operator should use the most appropriate means available under current plant conditions.
EOI Appendices 5A, 5B, and 6A provide guidance to operate CondensatelFeedwater, CRD, and only Condensate respectively. These systems are preferred sources of injection since they are of high quality water and are used for RPV water level control during normal plant power operations. Feedwater and CRD both provide high pressure injection from either a steam or motor-driven supply, and Condensate by itself provides for low pressure injection.
EOI Appendices 5C and 5D provide guidance to operate RCIC and HPCI respectively. The operator is instructed to operate RCIC and HPCI with suction from the CST if available, to ensure that the highest quality water is used for injection into the RPV. The CST is the preferred suction source not only because of higher water quality, but also because the CST is not subjected to the temperature increase that the suppression pool is. For these reasons, defeating HPCI high suppression pool suction transfer logic in EO! Appendix 5D, allows the operator to maintain the CST as the suction source. EOI Appendix 5C provides direction to defeat the RCIC low RPV pressure isolation interlock, that allows operation of the RCIC turbine at low pressure. Even if RPV pressure is below the isolation setpoint, but above turbine stall pressure, RCIC can still provide some injection into the RPV.
EO! Appendices 6B and 6C provide guidance to operate LPCI Systems I and II respectively. The operator is instructed to only operate RHR in LPCI mode when suppression pool level is above <A.62>. Engineering calculations have determined that operation ofRHR pumps below a suppression pool level of <A. 62> may induce vortex formation at the system suction strainer.
REVISION 0 PAGE 45 Of 110 SECTION O-V-K
C5, LEVEUPOWER CONTROL BASES EOI PROGRAM MANUAL SECTION ON-K . I DISCUSSION: STEP C5-30 I This signal step infonns the operator that actions to control RPV pressure control must immediately change because of present plant condition When emergency RPV depressurization is required, the operator shall transfer RPV pressure control actions from the RCIP Section of EO1-1, RPV Control, to C2, Emergency RPV Depressurizatio This step has been reached in this procedure because previous attempts to maintain adequate core cooling have been unsuccessful, or plant conditions are such that emergency RPV depressurization is required, as indicated by a signal step in another Eor being concurrently execute If adequate core cooling cannot be assured, then plant conditions may be such that RPV water level is at or below T AF, and RPV pressure is high enough to prevent injection from low-head pumps. Therefore, emergency RPV depressurization is required for the purpose of maximizing injection flow from high-head pumps and to permit injection from low-head pump Depressurizing the RPV is preferred over restoring RPV water level through the use of systems that inject inside the shroud because: A large reactor power excursion may result from the in-shroud injection of relatively cold wate . Rapid depressurization, by itself, will shut down the reactor due to a substantial increase in void . Following the depressurization, reactor power will stabilize at a lower level.
- REVISION 0 PAGE 81 OF 110 SECTION O*V-K
C5, lEVEUPOWER CONTROL BASES EOI PROGRAM MANUAL SECTION O-V-K I DISCUSSION: STEP C5-38 I Caution #5 applies throughout performance of Step C5-38. Caution #S is identified at this step to highlight the potential for large power excursions and subsequent core damage if cold, unborated water is rapidly injected using injection sources within Step C5-3 , This action step directs the operator to use injection sources listed to restore and maintain RPV water level above <A.71>. System specific EOl Appendices provide step-by-step guidance for lining up and injecting into the RPV. Injection pressures <<A.l>> have also been provided as additional information to the operato Engineering calculations have determined that when RPV water level is at or above <A. 71 >, adequate core cooling is still assured. The value of <A.71> RPV water level is Minimum Stearn Cooling RPV Water Level. Refer to discussion of Step C5-16 for more information on Minimum Stearn Cooling RPV Water Leve This step has been reached only when RPV water level cannot be restored and maintained above
<A.71> usmg preferred systems. Therefore, use of additional systems is required that either inject inside the core shroud, are difficult to lineup, or take suction on sources of comparatively lower 1 water quality. No priority between use of each listed system is intended, therefore, the operator should use the most appropriate means available under current plant condition EOI Appendices 6D and 6E provide guidance to operate CS Systems I and II. CS provides relatively high quality water from the suppression pool and can provide injection into the RPV quicker than other sources listed in this step. However, reactor power excursions are more probable since CS injects directly into the core shroud at high flowrates. Therefore, extreme caution should be used for CS injection at this ste Unlike directions given for use of motor driven pumps in EOI-l, RPV Control, CS System operation is not restricted by pump NPSH and Vortex limits (suppression pool level). Even though risk of equipment damage exists ifNPSH and Vortex limits are exceeded, immediate and catastrophic pump failure is not expected should operation beyond these limits be required. Since prolonged operation under these conditions is most likely required before degraded system and pump performance may result, the undesirable consequences of uncovering the reactor core outweigh risk of equipment damag EOI Appendices 7C, 7E, and 7F provide guidance to inject RHR into the RPV from crossties to other units or through RHR Drain Pumps A and B. EOI Appendix 7G provides guidance to inject into the RPVwith PSC Head Tank Pumps. All of these injection sources provide suppression
pool water at low pressure, but are relatively complicated to line up . REVISION 0 PAGE 101 OF 110 SECTION O-V-K
OPL 171.205 Revision 8 Page 60 of 73 Instructor Notes The systems listed in Step C5-13 for use in Cautions 2,3,5,6 controlling RPV water level comprise all those apply which inject outside of the core shrou These are used, preferentially, because the flow path outside the core shroud mixes the NPSH and Supp relatively cold injected water with the warmer PL Ivl limits rum.!Y water in the lower plenum prior to it reaching the cor The last bullet in the list of systems applies only if step C5-25 was executed for level control. Those systems listed in step C5-25 inject outside the shroud, or are of low quality water or are difficult to line up. If level restoration was performed using those systems, then it is appropriate to use them again when this step is reached.
12. Step C5-14 (Override) This override applies to steps C5-15 through C5-18 If while performing the associated steps, reactor power rises above 5% and RPV water level is above-50" direction to return to step C5-6 and re-evaluate conditions to determine a new water level band.
13. Step C5-15 This step is reached only if power was below 5%. There is no driving condition to lower level. The RPV water level band is -180" to
+5 As in step C5-13, those systems listed inject Cautions 2,3,5,6 outside the shrou apply NPSH and vortex limits rum.!Y 14. Step C5-16 (Override) Part One (1) Step C5-18, which follows, directs restoration of RPV water level to the normal range after having injected sufficient boron to shut down the reacto . RO 211 000AK2.0 1 00 lIe/AiT2G 1//63N.B.5/211 000AK2.0 112.9/3 .1IRO/SROIMODIFIED 11117/07 Given the following plant conditions: * Unit 1 is operating at 100% powe * A fire inside the board causes a loss of the '1 B' 480V Shutdown Boar * Fire Protection reports that the fire cannot be extinguishe * The Unit Supervisor (US) directs a manual scra * NOT ALL control rods insert, and the following conditions are noted: - Reactor Power 15% - Suppression Pool Temperature 108°F and rising * The 'A' 4kV Shutdown Board deenergized due to an electrical fault when '1A'
RHR pump was started for Suppression Pool cooling lAW EOI- Which ONE of the following describes the action and method of injecting boron into the reactor? Transfer the _ _ _ _......(1..:.....)<--_ _ _ _ to alternate and inject Standby Liquid Control (SLC) using the _ _ _..>..:(2=)_ __
(1 ) (2)
A'! '1A' 480V Shutdown Board; '1A' SLC Pum B. '1 B' 480V Shutdown Board; '1 B' SLC Pum C. '1A' 480V Reactor MOV Board; '1A' SLC Pum D. '1 B' 480V Reactor MOV Board; '1 B' SLC Pump.
Page: 1 of 2 2/19/2008
KIA Statement: 211000 SLC K2.01 - Knowledge of electrical power supplies to the following: SBLC pumps KIA Justification: This question satisfies the KIA statement by requiring the candidate to correctly identify the power supplies to the SLC pump References: OPL 171.039 rev 16, pg 15, 4.b & c 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 solve a problem. This requires mentally using this knowledge and its meaning to resolve the proble NRC Exam MODIFIED FROM 211 000K2.01 REFERENCE PROVIDED: None Plausibility Analysis: A - correct: '1A' and '1 B' SLC pumps are fed from their respective 480V Shutdown Boards. '1 B' is not available due to the loss of the '1 B' 480V Shutdown Board. The 'A' 4kV Shutdown Board feeds the '1A' 480V Shutdown Board. When the 'A' 4kV Shutdown Board is lost due to an electrical fault, the '1A' 480V Shutdown Board is de-energized. Manually transferring '1A' 480V Shutdown Board to Alternate will restore power to the board and allow starting '1A' SLC Pum B - incorrect: Due to the loss of 'A' 4kV Shutdown Board, the '1A' 480V Shutdown Board has lost power. This is plausible if the candidate is not aware that the 480V Shutdown Board does not automatically transfer to Alternate the same as the 4kV Shutdown Boar C - incorrect: The power supply to '1A' SLC is not '1A' 480V RMOV Board. It is plausible because '1A' 480V RMOV Board is safety related and powered from the same DG as '1A' 480V Shutdown Boar D - incorrect: The '1 B' 480V Shutdown Board is unavailable due to a fire. This answer is plausible if the operator does not know the correct power supply to SLC Pump Amplification: A fire in a 480V Shutdown Board may not result in it being lost depending on the location and severity of the fire. It is necessary to state the the '1 B' 480V SID Board is lost so that the candidate can assess the fact that he cannot transfer the board to alternate.
Page: 20f2 2119/2008
OPL 171.039 Revision 16 Page 15 of 48 INSTRUCTOR NOTES 4. SLC Pumps a) Two 100% capacity, triplex, positive displacement Obj. V.B. piston pumps are installed in paralle Obj. V.C Obj. V. b) 'A' pump is powered from 480V Shutdown Board Obj. V. c) 'B' pump is powered from 480V Shutdown Board Obj. V.B. d) Electrically interlocked so that only one pump will run at Obj. V. C. a time. This prevents system overpressurizatio Obj. V.B. Obj. V. C. e) The pumps are manually started from the main control room using the key-lock switch on panel 9-5, or locally, using the Test Permissive Transfer Switch at Panel 25-1 f) A control room start signal will fire the explosive valve A local start will not fire the explosive valve g) Either pump is capable of supplying a system flow of Obj. V.D. approximately 50 gpm at a system pressure of 1275 Obj. V.E. psi h) Each pump discharge has a relief valve, set at 1425 +/-. Obj. V.B. psig, to protect the pump and the system from Obj. V. C. overpressurizatio Obj. V.D. Obj. V.E. i) Each pump contains internal suction and discharge check valves, which open at approximately 5 psid, allowing only forward flow through an idle pump. (INPO O&MR 341).
j) Pump motors are protected by an undervoltage trip.
5. Accumulators a) An accumulator is installed between each pump and its discharge check valv b) Dampens the pressure pulsations that are inherent with Obj. V.D. piston-type, positive-displacement pump Obj. V.E. c) A steel vessel accumulator, containing a synthetic bladder, with one side charged to -450 psig nitrogen gas and SLC solution on the other sid KV SID
,.,
B t, C~,
-I
)NO l NO)
':'~J;w.]@,ffl.!'<\i\I",,~wr!'lI:!i:*,~i!!'J!<W>>!\-W~bWi.~~~.'!l'AA
~
I
.. V DSL Aux B A c:
$: I) 480V SID ~) NO NC ) 480V SID') NO NO NC Bd. 1A *
-1"1 .. ~.:w,~~t (1=
..a.
N ~Bn't'Wi' AM i w
.a::..
I) NC
~
NC ) ') NC NC) NC) !\J
~ ~ I) NC I I I ..,~
5 NO ) )NC NO) St.c...,
.Dv~P .,,1 ,I PillA ~.I .. _'" L c
r 480V OV RMOV Bd. 1A RMOV Bd. 2A i::: ) NO NO C"
~
o '1£1
- s
) NO NO-0>>;;00 fl) "U CD -0 cc"U< .....
A .f""l l j CD (1) Cij' ~ I\) ::J -' ""-.J
,V...I .....
_ . ::J 0>< ..... 0
- 0 ...... -....J W I\)
a
6. RO 212000K6.03 001IMEM/T2GIIRPSI1212000K6.03//RO/SROIBANK 11127/07 RMS Given the following plant conditions:
* Reactor Water Level Normal Control Range Instrument, 2-US-3-203A, has failed downscal Which ONE of the following describes the Analog Trip System Response?
The trip relay will be _ - - -(1 l . .). ! . J - - _ _ and the contact in the Reactor Protection System (RPS) Logic will be (2)
(1 ) (2) de-energized; closed 8. .... de-energized; open energized; closed energized; open KIA Statement:
212000 RPS K6.03 - Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR PROTECTION SYSTEM: Nuclear boiler instrumentation KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions involving level instrumentation to determine the response of RPS logic component References: OPL 171.003 rev 18, pg 28 (d), OPL 171.028 rev 17, pg 14,3.e(3) and 2-730E915 series print Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam Page: lof2 2/19/2008
REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Whether RPS relays are normally emergeized or de-energize . Whether RPS contacts fed from relays are normally open or close . Recall which scram signal, if any, is fed from 2-LlS-3-203 B - correct: RPS relays are normally energized and closed. When 2-LlS-3-203A fails downscale it is essentially the same as a low water level scram signal in that channe This will deenergize the associated relay which results in OPEN contact A - incorrect: This is plausible if the operator confuses PCIS logic relay response with RPS logic relay response and fails to recognize a valid trip has been generated by 2-LlS-3-203 C - incorrect: This is plausible if the operator fails to recognize a valid trip has been generated by 2-LlS-3-203 D - incorrect: This is plausible if the operator confuses PCIS logic relay response with RPS logic relay respons Amplification: This a MEMORY level question and the justification has been corrected.
Page: 2of2 2/19/2008
OPL 171.003 Revision 17 Page 54 of 54 t t >- ~C') <Il~ ~.Q ffl5"
~NSING -:;~Pjfft -----<rl (f) (f)
ffi '5 ~&!
~I "f' I I ril Jj'i;l II II Ll~l ~II~ cf 'T~ J I PI -1 9-6 H53 M [OJ ~ I I I~ l----~ : @~I@I~
BII~LLI
/
25-32 21>-32 G-5 G-5
/
3-~) / BAF2f
~/
9-5 135" JET PUMP f6 9-S LQll'ERTAP JI U i--\ ,w ' \'E'llj~ERO
:T ACCIDENT RANGE INSTR~ENTS (+32 TO .$.)
FUNCTION P11 LT3-62&S2 LT3-58 90&0 NI LT B0 & D Rx SCRMf POlS 236 & 6 VESSEL FLOODING RANGE 0 TO HPCI Rele Main & RFP TRI'S INST. NUl FUNCTI
. - - -- .- ._. - _.- * -_. -~.--. PT 3-22 C &. D PANEL 9-S INDICATIO PT3-7'-
TP-7: REACTOR VESSEL LEVEL/PRESSURE INSTRUMENTATION
OPL 171.003 Revision 17 Page 28 of 54 INSTRUCTOR NOTES (c) Also ensures that MCPR and LHGR are not violated during a feedwater controller maximum demand failure.
(2) High level alarm at +39 inches, Level 7 Defines upper end of normal operating region. If level is ~ 39 inches, transients such as a recirculation pump trip will not cause level to increase to turbine trip point.
(3) Low level alarm at + 27 inches, Level 4 Defines lower end of normal operating region. If level is ~ +27 inches, transients such as a reactor feedpump trip from full power will not cause level to decrease to reactor scram point with a runback of the recirculation pumps.
(4) Reactor scram at +2 inches, Level 3 (a) Assures the reactor will not be operated without sufficient water above the reactor cor (b) In conjunction with the containment isolation signal of +2 inches, prevents fuel cladding perforation (c) Prevents operation with separator skirts uncovered, which could result in carryunde (d) Set low enough to prevent spurious operation for normal operating transient (e) Starting SGTS establishes secondary containment.
(5) ATWS RPT/ARI trip at -45 inches, Level 2 Reduces core flow and bleeds off CRD HCU air header in event that low level was due to a failure to scra Also prevents operation of recirc pumps without adequate net positive suction hea OPL 171.028 Revision 17 Page 13 of 50 INSTRUCTOR NOTES (2) The third is used to produce manual SCRAM trip signals (trip channel A3).
(3) The channels for trip system Bare designated B1, B2 and B3. Both of the automatic channels in each trip system Obj. V.B. monitor critical reactor parameter Obj. V.DA (1) At least four channels for each monitored TP-3 parameter are required for the trip system Drawing logi E915RF-11 2-730E915RF-12 (2) If either of the two channels sense a parameter which exceeds a setpoint, then this would place the associated trip system (A or B) into a tripped conditio (3) To produce a SCRAM, both trip systems must be tripped. This is called a "one-out-or-two-taken twice" arrangement.
d. Each trip system logic may also be manually TP-4 trippe (1) Each .Trip system contains manual SCRAM Drawing switches on Panel 9-5 which cause a trip in 2-730E915RF-11 the respective trip system when actuate E915RF-12 (2) The Reactor mode switch has contacts in both the A3 and B3 channels. Placing the reactor mode switch in SHUTDOWN will result in a trip of both trip system (3) A trip in both channels A3 and B3 initiates a reactor SCRAM.
e. DUring normal operation SER 3-05 Operator fundamental (1) All sensor and trip contacts essential to safety are close (2) Channels, logics, and actuators are energize OPL 171.028 Revision 17 Page 14 of 50 INSTRUCTOR NOTES (3) WhenaSCRAMsignal is received, the logic Drawing relays deenergize to cause a SCRA E915-13 (4) Loss of power to one RPS bus will result in a Obj. V. half-SCRAM. Loss of power to both RPS Obj. V. buses will result in a full SCRA Obj. V.D.8 4. Reactor SCRAM Signals and Arrangement Obj. V. Refer to 01-99 for the setpoints for each SCRA Obj. V.CA SER 3-05 Operator Channel test switch fundamentals (1) Allows for testing each channel's trip functio Drawing 2-730E915RF-11 2-730E915RF-12 (2) Four, one per channel located on Panel 9-15 REACTIVITY and 9-17 in Aux. Inst. Roo MANAGMENT Discuss when switches (3) Key-locked, two positions - NORMAL and can be use TRIP (4) TRIP de-energizes that channel's relays producing a half-SCRA Turbine Stop Valves, 10 percent closure TP-5 anticipates the pressure and neutron flux rise Drawing caused by the rapid closure of the Turbine Stop 2-730E915-9,10 Valve Obj. V. (1) Each of the four Turbine Stop Valves is equipped with two limit switches. One limit switch is assigned to RPS "A" and one to RPS "B".
(2) These switches will provide a valve-closed signal to the RPS trip logi (3) The position switch contacts are arranged so Drawing that any two Stop Valves can be closed 2-730E915RF-11 causing no more than a half-SCRA E915RF-12 (4) Closure << 90% full open) of any combination of three Stop Valves will cause a full SCRAM in all case OPL 171.017 Revision 13 Page 12 of 56 INSTRUCTOR NOTES C. Typical PCIS Isolation Logic TP-1 A typical logic arrangement for the PCIS valves PCIS de-energizes (except MSIVs) is shown in TP-1. This figure shows to isolate (except that two separate trip channels (A and B) are each HPCIIRCIC) provided with two sensor relay contacts (A/C and BID). Obj. V. Obj. V. This arrangement creates trip subchannels A 1/A2 and B1/B A trip of either sensor relay within a trip HPCIIRCIC are channel will cause opening of the associated energize to actuate contact and de-energization of the associated relay. This condition will create a Obj. V. "half isolation" signal within both logic Obj. V. channels but NO VALVE MOVEMEN Should a trip of either sensor relay in the other trip channel occur, conditions will exist to de-energize the valve actuation relays in each logic channel, causing both isolation valves to clos PCISlogic is arranged as follows: A10RA2 AND = Inboard AND Outboard valve closure B1 OR B2 Note: Most PCIS logic is assembled as abov The MSL drains however are an exceptio The MSL drain logic is as follows: A1 AND B1 = liB valve closure A2 AND B2 =O/B valve closure
7. RO 215003A4.04 00IIMEM/SYS/IRMIB6/215003A4.04//RO/SROIMODIFIED 11117/07 Given the following plant conditions:
* Unit 3 reactor startup preparations are in progress with NO rods withdraw * Instrument Mechanics are performing the Intermediate Range Monitor (IRM)
functional surveillanc * No IRMs are currently bypasse * The Instrument Mechanic technician has placed the "INOP INHIBIT" toggle switch for the 'H' Channel IRM in the "INHIBIT" positio Which ONE of the following describes the IRM trip function that is bypassed as a result of this action? IRM "High Voltage Low" INOP TRIP is bypasse IRM "Loss of.+/- 24 VDC" INOP TRIP is bypasse IRM "Module Unplugged" INOP TRIP is bypasse D." IRM "Mode Switch Out of Operate" INOP TRIP is bypasse KIA Statement: 2150031RM A4.04 - Ability to manually operate and/or monitor in the control room: IRM back panel switches, meters, and indicating lights KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific component manipulations to correctly determine the response of the IRMsyste References: OPL 171.020 rev. 10, pg 22, c.(1) & (2) and 3-01-92A rev. 14, pg 7 of 15 P&L 3. Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam MODIFIED FROM OPL 171.020 #28 Page: lof2 2119/2008
REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. The function switch is bypassed by the "INOP INHIBIT" pushbutto NOTE: Each of the possible answers below will typically initiate an INOP trip of it's associated IRM channel, therefore each distractor is plausibl D - correct: INOP/INHIBIT Pushbuttons (toggle sw for Unit 3) are pushed/flipped to bypass the INOP trip that results from taking mode switch S-1 out of "OPERATE."
They are used to allow testing of other scram or rod block signals from the IRM drawer into RPS/RMCS without them being masked by the INOP tri A - incorrect: This INOP trip will still functio B - incorrect: This INOP trip will still functio C - incorrect: This INOP trip will still function.
( Page: 2of2 2119/2008
OPL 171.020 Revision 10 Page 20 of 42 INSTRUCTOR NOTES Trips TP-10 Rod blocks ObjV.D.7, V. Obj. V.C.3., Block Setpoint When Bypassed Downscale .:: Range 1 or RUN
> 90 RUN Mode Obj. V. Obj. V.CA.
INOP -HV low <<90v) RUN Mode Obj. V. Module unplugged-Function switch not in OPERATE-Loss of +/-24VDC Detector Wrong Detector RUN Mode ObjV.B.13 Position Not Full IN Scrams TP-11 Scrams Setpoint When Bypassed Obj. V. Obj. V.C.5. ObjV.D.8 High-High > 116A In RUN Mode INOP -HV low <<90v) In RUN Mode-Module unplugged-Function switch not in OPERATE-Loss of +24VDC Controls Provided Panel 9-5 Recorder switches switch between IRM channels and APRM/RBM channels Range switches allow operator to select appropriate IRM range to maintain indications between 25 to 75 on 0-125 scale. 0-40 scale is no longer utilize OPL 171.020 Revision 10 Page 22 of 42 INSTRUCTOR NOTES (2) 'Standby' - same as operate, except gives Inop trip to yield maximum design protection before channel is removed from servic (3) 'Zero l' - Removes signal from output amplifier so that output amplifier, local meter and recorder can be zeroe (4) 'Zero 2' - Removes voltage from range switch. This deselects all range This, in turn, causes no input to be sent to attenuator and allows setting the zero adjust on output amplifie (5) '125 '- Input is removed from attenuator same as Zero 2 position. A calibration signal is substituted which will yield 125 on the 125 scale. Used to set gain of output amplifie (6) '40' - Produces a 40 reading on the 125 scale.
c. INOPIINHIBIT Pushbuttons Obj. V. Obj. V.CA (1) Pushed to bypass the INOP trip that results from taking mode switch S-1 out of DCN W18726A
"operate." replaced the INOPIINHIBIT (2) Used to allow testing of other scram or rod Pushbutton with a block signals from the IRM drawer into toggle switch for the RPS/RMCS without them being masked by U-3 IRM drawer the INOP tri (UNIT DIFFERENCE)
0610 NRC RO EXAM 8. RO 215004A3.03 001/CIA/T2G1ISRMlB81215004A3.03//RO/SROIBANK 11127/07 RMS Given the following plant conditions:
* A reactor startup is in progress following refueling, with all Reactor Protection System (RPS) Shorting Links remove * The reactor is approaching criticalit * An electronic failure in the 'B' Source Range Monitor (SRM) drawer results in an SRM HIGH/HIGH output signa Which ONE of the following describes the plant response? A Rod Out Block ONL A Rod Out Block and 1/2 Scra A "SRM HIGH/HIGH" alarm ONL D." A Full Reactor Scra KIA Statement:
215004 Source Range Monitor A3.03 - Ability to monitor automatic operations of the SOURCE RANGE MONITOR (SRM) SYSTEM including: RPS status KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to correctly determine the response of the SRM with the shorting links remove References: 1~OI-92 rev. 6 pg 6 of 14, P&L 3.0.G, OPL 171.019 rev 12, pg 23, g(1) & pg 24 g(2), OPL 171.028 rev 17 TP-4 and OPL 171.148 TP-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 outcom NRC Exam Friday, February 29,20085:18:39 AM 16
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. The SRM response to a Hi/Hi output signa . The SRM response to the same conditions with the shorting links remove D - correct: The Reactor Protection System, in conjunction with the Neutron Monitoring System (SRM and IRM), has non-coincident trip logic IF all eight shorting links are removed. In this condition the SRM Hi-Hi will generate a scram. In addition, a Rod Out Block is also generated but will not be relevent following a full scra B - incorrect: This is plausible if the operator determines a scram signal is generated with the typical "1-out of-2 taken twice" logic. However, if all eight shorting links are removed, this condition will generate a full scra C - incorrect: In this condition the SRM Hi-Hi will generate a scra A - incorrect: This is plausible if the operator determines a scram signal is NOT generated with the shorting links removed.
( Friday, February 29,20085:18:39 AM 17
-I '"C I ...: LOCAL REACTOR CONTAINMENT Panol9-5 en l
LCR METER o c S1
- 0 (")
m o ZERO LCRTRIP TEST sg 81 PERIOD z
~
o 10 G) o 105 m
~
PERIOD TRIP S DRIVE o SH~FT 1 0-- Z FIXED -0 VARIABLE-I WALL RAMP o
~
2:! LOCAL z PERIOD AMP
>I PERIOD METER G)
(")
- c
>>z REF DP z m r .1 C Z (")
+20VDC + 15VDC
- ! MODULE o
z REF >> r OJ r +/- 24vdc o(")
~ ""0>>:;00 c SCRAM WITH SHORTING LINKS REMOVED ru "0 CD ""0 (0"0<.- >>
G) CD CD
:::l CIi"
-"--J
~
~3:g-"
~ OX-"O s -ON....\.
.t>. CD CD
OPL 171.028 Revision 17 Appendix C Page 41 of 50 ShortinQ Link (Remove both for I 5A-K13A (13B) non-coincidence SA-K13C NMS SA-K1SA SA-K1SC neutron monitor (13D) protectio NON-Remove on11.1 red COINCIDENT for coincidence neutron monitor protection .) UJ
::: ...J III I SA-K13B (13A)
TRIP ("B" CHANNEL IN PARENTHESES) SA-K13D (13C) Closes with MS in 12011 * 60HZ * 1a Shutdown From RPS Bus "A" SA-K16A Open in shutdo .... n _______ _ SA-S1 position 5A-17A TDC R 5A-K19A 5A-K19C RESET RESET TP-4: RPS MANUAL TRIP CHANNEL A3 (TYPICAL FOR 83)
OPL 171.148 Revision 8 Page 139 of 150 Contacts open on Contacts open on respective trip respective trip intwo or more in two or more Contacts close on non bypassed Contacts close on non bypassed bypass selection APRMs bypass relection APRMs I I I I I Closed In I I Run mode t I I l,.
,
Bypass igh High I
, ,
,,
'K1Y HIGH/INOP
" ,-h CHANNELA1 SCRAM K4Y (PRM HIG-t 5A-K12A SUBCHANNEL NEUTRON tv1ONITORING ISA-K12E SA-K12A SA-K12E SA-K13A NMS Non Coincident Trip TP-10 TYPICAL RPS CONFIGURATION FOR NEUTRON MONITORING SYSTEM
BFN Source Range Monitors 1-01-92 Unit 1 Rev. 0006 Page 6 of 14 PRECAUTIONS AND LIMITATIONS To prevent a rod withdrawal block when withdrawing SRMs, SRM count rate is required to be above Retract Permit (145 counts per second) or all un bypassed IRM channels are set to Range 3 or above and indicating above their downscale trip point (7.5 on 125 scale). Only one SRM channel can be bypassed at a tim In order to prevent an inadvertent rod withdrawal block or Reactor scram (with shorting links removed) while operating the SRM BYPASS selector switch, 1-HS-92-7A1S3,
* Verify the previously bypassed channel returns to normal status by observing the applicable HIGH HIGH and HIGH or INOP status lights are extinguished prior to selecting any other channel to be bypasse * After bypassing a channel, the applicable BYPASSED status light should be illuminated prior to testing, operating, or working on that channe To prevent SRM detector drive damage, the CRD service platform should be locked in the stored position with key removed to allow free movement of SRM In order to minimize their exposure, SRM detectors should be fully withdrawn from the core when IRMs are on range 3 or above and indicating above their downscale trip poin Illustration 1 lists trip signals and associated actions for the Source Range Monitoring Syste The Reactor Protection System in conjunction with the Neutron Monitoring System (SRM and IRM) has non-coincident trip logic if all eight shorting links are removed. If only the yellow, green, and red shorting links (six total) are removed, the SRM High-High trips will be placed in a one-out-of-two taken twice logi The time required to drive a detector from full-out tofull-in is approximately 3 minute The INOP/INHIBIT switches, located on Panel 1-9-12 SRM drawers, bypass the SRM switch position out-of-operate trip. These switches are to be used only during testing of SRM channel [NRC/C] Upon return to service of 24 VDC Neutron Monitoring Battery A or B, Instrument Maintenance is required to perform functional tests on SRMs and IRMsthat are powered from the affected battery board. [NRC IE Inspector Foliowup Item 86-40-03]
9. RO 215005A2.03 001lCIAlT2G1IPRNM/APRMlB7/215005A2.03//RO/SROIBANK 11127107 RMS Which ONE of the following describes the expected response due to a "FAULTil in an Average Power Range Monitor (APRM) channel and the required action(s) to address this condition? An APRM channel _ _---'-(1~)_ _ _ will result in an INOP trip to_ _ _~(2::....)_ __ To continue plant operation, bypassing the APRM _---'('""'3~)___ require (1 ) (2) (3) Critical Fault; ONL Y the respective 2/4 logic module voter; is NOT Non-critical Fault; ONLYthe respective 2/4 logic module voter; is NOT Non-critical Fault; ALL four of the 2/4 logic modules voters; is D.oI Critical Fault; ALL four of the 2/4 logic modules voters; is KIA Statement: 215005 APRM I LPRM A2.03 - Ability to (a) predict the impacts of the following on the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions Inoperative trip (all causes).
KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to correctly determine equipment response and the corrective actions due to an APRM INOP tri References: 01-92B Precautions and Limitations 3.0.1, 1-ARP-9-5A rev. 11 window 25 and OPL171.148 rev 8, pg 3 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 outcom NRC Exam Page: 10f2 2/19/2008
REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. The difference between a "Critical Fault" and "Non-critical Fault" with respect to PRNM respons . The operation of the 2/4 Logic Module Voter operation for each type of faul . Based on the above, determine the appropriate course of action regarding the APRM channel in questio D - correct: "Critical Faults" are those that affect the instrument's capability to perform its intended function and will cause an instrument INap trip to all four 2/4 logic module voters and a trouble Alarm indication. Per the ARP it is allowable to bypass the APRM since no other APRMs are bypasse A - incorrect: This is plausible because a "Critical Fault" generates an INap trip input to the 2/4 Logic Module Voters, but to all four modules, not just two. In addition, not placing the APRM in BYPASS because ofa "Critical Fault" is inappropriate since any additional equipment failure could result in an unnecessary scra B - incorrect: A "Non-critical Fault" will not result in an INap trip input. This is plausible because a "Non-critical Fault" generates a Trouble Alarm similar to a "Critical Fault". In addition, not placing the APRM in BYPASS because of a "Non-critical Fault" is appropriat C - incorrect: A "Non-critical Fault" will not result in an INap trip input. This is plausible because a "Non-critical Fault" generates a Trouble Alarm similar to a "Critical Fault". In addition, placing the APRM in BYPASS would be correct IF an INap trip was generate Amplification: will need clarification on you're "true/false" concern? Page: 2of2 2/19/2008
BFN Panel 9-5 1-ARP-9-5A Unit 1 1-XA-55-5A Rev. 0011 ( Page 31 of 43 Sensor/Trip Point: APRM HIGHIINOP APRM: C. HIG OROPRM TRIP 1. (0.66W + 65%) Two Loop operation with Rx MODE Switch in Ru . 14.0% with Rx Mode Switch not in RU (Page 1 of 1) 3. 119% in any mod . (0.66(W-10%)+65%) Single loop operation with Mode Switch in RU D. INO . APRM Chassis Mode switch not in OPERAT . Loss of input powe . Watchdog timer timed ou . Critical self test fault detecte OPRM TRIP: Anyone of the three algorithms, period, growth, or amplitude for an operable OPRM Cell has exceeded its trip value condition Sensor Control Room Panel 1-9-1 Location: Probable Flux level at or above setpoin Cause: Testing in progres Malfunction of senso Control rod drop acciden Thermal-Hydraulic instability detected Automatic A. Input signal to all four Voters, no output to RP Action: B. Full Reactor scram if two sensors actuat Operator A. VALIDATE alarm by multiple indications D Action: B. With SRO permission, BYPASS initiating channel to reset the alar REFER TO 1-01-92 D C. IF HIGH is indicated, THEN DETERMINE cause of inadvertent reactivity chang D Continued on Next Page
OPL 171.148 Revision 8 Page 31 of 150 b. Each LPRM instrument provides a brief Obj, V,D,4 description of the self-test faults which are divided into two categories, "Critical" and
"Non-Critical" fault (1 ) Critical faults are those that affect V.B.? V. the instrument's capability to perform The Trouble its intended function and will cause Alarm is indicated an instrument INOP trip and a in the Status Trouble Alarm indicatio Header for each instrumen (2) Non-critical faults do not prevent the instrument from performing its intended function and will cause a Trouble Alarm indication only.
c. The LPRM instrument transmits its self-test status to its associated APRM and RBM instrument BFN Average Power Range Monitoring 1-01-92B Unit 1 Rev.OOOa Page 7 of 27 PRECAUTIONS AND LIMITATIONS (continued) Each of the four APRM/OPRM channels input to the four Voters, such that when a signal is generated from an APRM/OPRM channel, all four Voters see and reflect that signal. Each Voter is directly associated with one RPS sub-channe When operating in a 2 out of 4 voting configuration, the first un-bypassed input will be seen as a single input with no trip outputs. When the second un-bypassed signal of the same type [The SAME TYPE inferring that one type is an APRM function and a different type is an OPRM function] is received it will also be seen by all four Voters resulting in a trip output from all four Voters consequently producing a full reactor scra Bypassing an APRM does not preclude testing a Voter, such that with an APRM in bypass, the Voters can still be tested and produce half scrams. Voters are not bypassed with the APRM joystic The Recirc Flow Indication and the Voters are never bypassed unless they are removed for testing. There is no bypass capability for the Recirc flow signal input or Voter A reactor scram will be produced when at least two of the SAME TYPE of trip inputs are received by the Voters: Either: APRM HIGH/INOP {i.e., APRM High Flux/STP Flow Biased Scram/INOP} OR: Any OPRM ABA, PBA, or GRBA algorithm trip conditions me The SAME TYPE inferring that one type is an APRM function and a different type is an OPRM functio The new APRM modules contain an automatic power oscillation detection and suppression function (Oscillation Eower Range Monitoring) which detects and protects against thermal hydraulic instabilities. OPRM monitors local cell area for thermal hydraulic core instabilities. There are 4 channels each containing 33 cells. Each cell contains up to 4 LPRM inputs per OPRM channel for power monitorin Oscillations are detected using anyone of three algorithms; Period Based Algorithm, Growth Rate Based Algorithm, and, Amplitude Based Algorith When power oscillations are detected a trip signal inputs to the Voters which will in turn, send a trip output to the RPS sub-channels and will produce a trip Signal. Two of these types of signals will produce a full reactor scra BFN Average Power Range Monitoring 1-01-92B Unit 1 Rev.OOOS Page 19 of 27 Illustration 1 (Page 1 of 6) APRM/OPRM Trip Outputs and PRNMS Overview APRM Trip Outputs TRIP SIGNAL SETPOINT ACTION APRM Downscale 5% 1. Rod Block if REACTOR MODE SWITCH in RUN.
APRM Inop APRM Chassis Mode not in 1. One Channel detected, no alarm or OPERATE (keylock to INOP). RPS output signa . Loss of Input Power to APR . Two Channels detected, RPS output signal to all four Voters (Full Reactor Scram). Self Test detected Critical Fault in the APRM instrumen . Firmware Watchdog timer has timed out.
APRMlnop < 20 LPRMs in OPERATE, or 1. <20 LPRMs total or <3 per level results Condition < 3 LPRMs per leve in a Rod Block and a trouble alarm on the display panel. This does not yield an automatic APRM trip, but does, however, make the associated APRM INOP.
APRM High DLO 1. Rod Block if REACTOR s;; (0.66W + 59%) MODE SWITCH in RU SLO s;; (0.66 (W-liW) + 59%)
[W = Total Recirc drive flow in %
rated]. Neutron Flux Clamp Rod Block 2. Rod Block if REACTOR s;; 113% MODE SWITCH in RU . s;; 10% APRM Flu . Rod Block in all REACTOR MODE SWITCH positions except RUN.
APRM High High . Scram DLO s;; (0.66W + 65%) SLO s;; (0.66(W-I1W) + 65%)
[W,= Total Recirc drive in %
rated]. s;; 119% APRM FLU . s;; 14% APRM FLU . Scram in all REACTOR MODE SWITCH positions except RUN.
Recirc Flow s;; 5% mismatch between APRM 1. Flow compare inverse video alarm.
Compare Channels.
Recirc Flow % Flow Monitor upscal . Rod Block.
Upscale
10. RO 217000K2.03 001lC/AJT2G 1IRCIC/7/217000K2.03/IRO/SROIBANK 11127/07 RMS Given the following Unit 2 conditions:
* The Control Room has been evacuate * Reactor Core Isolation Cooling (RCIC) is controlling reactor water leve * A loss of the Division I Emergency Core Cooling System Inverter (Div 1 ECCS Inverter) occur Assuming no further operator action, Which ONE of the following describes the RCIC system response?
The RCIC Flow Controller_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ lowers to minimum in auto ONL raises to maximum in manual ONL C.oI lowers to minimum in either manual OR auto mod raises to maximum in either manual OR auto mod KIA Statement: 217000 RCIC K2.03 - Knowledge of electrical power supplies to the following: RCIC flow controller KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to correctly determine a loss of the logic power supply to the controller has occurred and the RCIC system response to that loss .. References: 2-01-71 Precautions and Limitations 3.0.W, OPL 171.040 rev 22, pg 34, 8.a(1) & (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 outcom NRC Exam Page: 10f2 2/19/2008
REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. The current power supply to the RCIC controller while being operated from the Backup Control Panel, 25-3 . The Power Transfer Switch (XS-256-1) is NOT part of the RCIC initiation procedure in 2-AOI-100-2, "Control Room Abandonment".
3. The failure mode of the Yokogawa Flow Controller used for RCIC while operating from the Backup Control Panel, 25-3 NOTE: Due to the widespread use, at BFN, and various failure modes of Yokogawa Flow Controllers, each of the four answers become plausible. These controllers can be set to fail "as-is", "fail high" or "fail low" depending on the system and applicatio C - correct: Loss of Power to the Flow Controller (Div I ECCS ATU Inverter) results in the controller output dropping to zero milliamps and turbine speed would lower to minimum (600 rpm). However, on Unit 3, the Div I ECCS Inverter also powers to EGM Control Box which would result in overspeed on Unit 3 ONL A - incorrect: The RCIC controller at Panel 25-32 fails low in either AUTO or MANUA B - incorrect: The RCIC controller at Panel 25-32 fails lo D - incorrect: This is plausible due to the effect it would have on Unit 3, versus Unit Amplification: As stated in P&L 'W', the power supply was changed. It did change the answer in the original question; so the stem was modified to include the correct power supply. This is considered a BANK question because none of the distractors were modified.
Page: 20f2 2/19/2008
OPL 171.040 Revision 22 Page 21 of 74 INSTRUCTOR NOTES RCIC PUMP PI-71-20A 0-50 psig SUCTION PRESSURE RCIC TURBINE PI-71-4A 0-1500 psig STEAM LINE PRESSURE RCIC TURBINE PI-71-12A 0-50 psig BFPER971133 EXHAUST Indicator could read PRESSURE from 0-200 rpm in standby RCIC TURBINE SI-71-42A 0-6000 rpm readiness due to SPEED non-linearity in low RCIC TURBINE FI-71-1A 0-80 Ibm x1000 RPM range STEAM FLOW FI-71-1 B 2. Flow Controller (FIC-71-36A & B) a. One located on Panel 9-3 and one on Panel 25-32 (Remote Shutdown Panel).
b. Power Supply to the Pnl 9-3 controller (FIC-71- Obj. V. A) is the Div I ECCS Inverter c. Power Supply to the Pnl 25-32 Controller (FIC-71-36B) is also the Div I ECCS Inverte d. AT Pnl 25-32, there is a power transfer switch TP-13 (XS-256-1) which, if placed in the Alternate position, will transfer both (36A & B) Flow Controller power supplies from the Div I ECCS Inverter to the Unit Preferred 120VAC Power Supply.
3. Yokogawa Flow Controller a. AUTO - output signal is changed by changing the setpoint. Full Scale travel of setpoint is 40 seconds. Momentary depressing of either the raise or lower keys will cause ~0.7 gpm change (~1 %).
OPL 171.040 Revision 22 Page 34 of 74 INSTRUCTOR NOTES 8. Failure Modes a. Loss of Power to the Flow Controller Obj. V. (1 ) Div I ECCS ATU Inverter (2) Loss of Power causes the controller to go to zero Reference P&L milliamp output and turbine speed would lower to 3.23/3. minimum (~600 rpm). However, on U3, the Div I ECCS Inverter also powers to EGM Control Box which would result in overspeed on Unit 3 onl b. Loss of control air ObjV. Obj. V. (1 ) RCIC steam line steam trap bypass valve (FCV-71-5) fails closed (Unit 3)
(2) RCIC steam line condensate drain valves (FCV 71-6A and 6B) fail closed (3) RCIC condensate pump Clean Radwaste discharge valves (FCV-71-7A and 7B) fail closed c. Loss of electrical power to valves Obj. V. Obj. V. All motor-operated isolation valves remain in the last position upon failure of valve power. Solenoid operated valve FCV 71-5 (Unit 2) fails close d. Loss of Power to Relay Logic Obj. V. Obj. V. (1 ) If Bus A fails, the automatic initiation circuit and UNIT turbine trip solenoid will not operate. Channel A DIFFERENCE isolation logic circuit is lost. Power is lost to EG-M control box and this causes FCV-71-10 trip governor valve to go wide open (if RCIC is operating). - Unit 2 (Unit 3 EGM power is from DIV I Inverter) (2) If power is lost to the EGM Control Box, Springs UNIT will re-position the 71-10 servo to fully open the DIFFERENCE governor valve (Unit 2 only).
BFN Control Room Abandonment 2-AOI-100-2 Unit 2 Rev. 0051 ( Page 11 of 95 Unit 2 Subsequent Actions (continued) NOTES 1) Attachment 1 provides normal backup control stations and available communication ) Attachment 10 provides PAX extensions and location [7] ESTABLISH communication with the following personnel and DIRECT attachments be completed as follows:
* U-2 Unit Operator complete Attachment 2, Part D * U-2 Rx Bldg AUO complete Attachment 3, Part D * U-2 Turb Bldg AUO complete Attachment 4, Part D CAUTION RCIC TURBINE STEAM SUPPLY VALVE, 2-FCV-71-8, transfer switch has been placed in EMERGENCY and will NOT trip on Reactor Water Level High (+51 inches). Failure to maintain level below this value may result in equipment damag RCIC will still trip on low suction pressure, high turbine exhaust pressure, mechanical overspeed, and trip push button on pnl 25-3 [8] Upon completion of attachments, RE-ESTABLISH communication using the best available means and continue procedur _ 1 ~ D 7<1 ND e&CWI t2eMc-N T TO /MN~F61 [9] INITIATE RCICasfoliows: Rc.tc... Lo~TJZ.GLL~ PDwel2- :;:'OffLj' [9.1] At Panel 2-25-32, CHECK OPEN 2-FCV-71-9 (Red Light above switch) RCIC TURB TRIP/THROT VALVE RESET, 2-HS-71-9 D [9.2] At 250V DC RMOV Bd 2B, compt. 50, PLACE RCIC PUMP MIN FLOW VALVE EMER HAND SWITCH, 2-HS-071-0034C, in OPEN. (Unit 2 Turbine Building AUO) D [9.3] At 250V DC RMOV Bd 2C, compt. 4B, PLACE RCIC TURB STM SUPPLY VALVE EMER HAND SWITCH, 2-HS-071-0008C, in OPEN. (Unit 2 Reactor Building AUO) D
BFN Control Room Abandonment 2-AOI-100-2 Unit 2 Rev. 0051 Page 12 of 95 Unit 2 Subsequent Actions (continued) NOTE RCIC Turbine should start and flow should stabilize at 600 gp [9.4] At Panel 2-2S-32, CHECK turbine speed 2100 rpm or above using RCIC TURBINE SPEED, 2-SI-71-42 o
[9.S] At 2S0V DC RMOV Bd 2B, compt. SO, PLACE RCIC PUMP MIN FLOW VALVE EMER HAND SWITCH, 2-HS-071-0034C, in CLOSE. (Unit 2 Turbine Building AUO) o [9.6] At Panel 2-2S-32, ADJUST flowrate as necessary using RCIC SYSTEM FLOW/CONTROL, 2-FIC-71-36 D [9.7] At Panel 2-2S-32, MAINTAIN Reactor Water Level between +2 and +SO inches using RXWATER LEVEL A & B, 2-Ll-3-46A & D NOTE The following step prevents HPCI operation and automatic opening of HPCI MAIN PUMP MINIMUM FLOW VALVE, 2-FCV-73-3 [10] At 2S0V Reactor MOV Bd 2A, PERFORM the following: [10.1] Compt. 3D, VERIFY CLOSED HPCI STEAM SUPPLY VALVE TO TURB FCV-73-16 (MO 23-14). D [10.2] Compt. 3D, PLACE HPCI TURBINE STEAM SUP VLV TRANS, 2-XS-73-16, in EMER D [10.3] IF desired to verify HPCI MIN FLOW BYPASS TO SUPPRESSION CHAMBER VALVE, 2-FCV-73-30, closed prior to opening breaker, THEN DIRECT operator to verify locall D [10.4] Compt. 80, PLACE HPCI MAIN PUMP MIN FLOWVLV FCV-73-30, breaker in OF D
BFN Reactor Core Isolation Cooling 2-01-71 Unit 2 Rev. DOSS Page 11 of 70 PRECAUTIONS AND LIMITATIONS (continued) Suppression pool water temperature should not exceed 95°F without suppression pool cooling in service to restore temperature to less than or equal to 95°F within 24 hour RCIC Testing is NOT permitted with suppression pool water temperature above 105° After RCIC steam lines have been hydrostatically tested, leak tested, or exposed to other conditions which could fill the 2-FCV-71-2 valve bonnet with water, 2-FCV-71-2 should be cycled to prevent overpressurizatio [II/FJ Prior to initiating any event which adds, or has the potential to add, heat energy to the suppression chamber, the Unit Supervisor will evaluate the necessity of placing suppression pool cooling in service. This is due to the potential of developing thermal stagnation during sustained heat addition [11-8-91-129J Calculations have shown that 16 min. of RCIC operation without RHR operating in the Suppression Pool Cooling Mode will result in a one deg F rise in bulk suppression pool temperatur [NER/C] Extended RCIC System operation may raise suppression chamber O 2 concentration above TRM 3.6.2 limits because of air-inleakage from RCIC Turbine Gland Seal System. [GE SIL 548J Whenever the 1 E ECCS ATU Inverter (Division I) becomes INOP, RCIC Is considered INOP. DCN W17726B changed power supply for RCIC flow controller from 1 E Unit Preferred MMG set busses to the Unit 2 1 E ECCS ATU Inverter (Division I). The RCIC STEAM LINE OUTBD ISOLATION VLV hand switch, 2-HS-71-3A, must be held in the OPEN position until 2-FCV-71-3 is fully open because the open seal-in circuit has been removed per ECN P016 [INPO/C] A buildup of corrosion products in the RCIC TURBINE CONTROL VALVE stem packing could result in speed oscillations, failure to control at the desired speed, and mechanical overspeed of the RCIC Turbine. During operation, RCIC Turbine parameters such as time to reach operating speed, speed stability, and governor response should be monitored to identify possible corrosion product buildup in the RCIC TURBINE CONTROL VALVE. [INPO SER 95004J (11/C)During routine plant evolutions, notify RADCON prior to making changes in the RCIC System which could cause a rise in area radiation levels. Confirm RADCON has implemented appropriate radiological controls/barriers for the expected RCIC System alignment prior to performing the alignmen (BFPER961778)
11. RO 218000Kl.05 001IMEMlT2G1I100-2/5/218000Kl.05//RO/SROIBANK 11127107 RMS Given the following plant conditions:
* The Unit 1 and 2 Control Rooms have been abandone * ALL Panel 25-32 Safety Relief Valve (SRV) Transfer Switches have been placed in EMERGENC * ALL Panel 25-32 SRV Control Switches have been verified in CLOS Which ONE of the following describes the operation of the SRVs associated with the Automatic Depressurization System (ADS)?
The associated ADS valves will OPEN _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ upon receipt of an ADS initiation signa B.oI in Safety mode, if their respective pressure setpoints are exceede in Relief mode, if reactor pressure exceeds the relief mode setpoin if their respective control switches, on Panel 9-3, are placed in OPE KIA Statement: 218000 ADS K1.05 - Knowledge of the physical connections and/or cause- effect relationships between AUTOMATIC DEPRESSURIZATION SYSTEM and the following: Remote shutdown system: Plant-Specific KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to correctly determine their effect on MSRV operation during a Remote Shutdown conditio References: OPL 171.208 Rev. 5 page 8, 2-AOI-100-2 page 8 and OPL 171.009 rev 10 TP- Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam Page: lof2 2/19/2008
REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Transferring the SRV control to Panel 25-32 disables the ADS functio . Transferring the SRV control to Panel 25-32 disables the Panel 9-3 control switc . Transferring the SRV control to Panel 25-32 does NOT disable the Pressure Relief functio B - correct: Transferring the SRV control to Panel 25-32 disables the ADS function and the Panel 9-3 control switch. The SRV's will still operate in safety mode but not relief mode, ie. the pressure switches that input into the logic are bypassed when control is transferred to panel 25-3 A - incorrect: Transferring the SRV control to Panel 25-32 disables the ADS functio C - incorrect: The pressure switches that input into the releif mode logic are bypassed when control is transfered to panel 25-3 D - incorrect: Transferring the SRV control to Panel 25-32 disables the Panel 9-3 control switc Amplification: Changed distractor 'C' ('D' from previous NRC comments) to make it more plausible.
Page: 2of2 2/19/2008
OPL 171.20S Revision 5 Page S of 30 INSTRUCTOR NOTES Trip reactor feed pumps as necessary to prevent tripping Obj. V. on high water leve Obj. V. . Start the diesel generators. (9-S Switch starts respective units DIG only) 1 Verify each EECW header has one pump in servic . Announce to all plant personnel that the Control Room is being evacuated and all operators are to report to their assigned backup control station . Obtain hand held radios from the control roo . Proceed to the Backup Control Panel (25-32) F. Subsequent Actions See AOI-1 00-2 for details for actions HU Tools: Procedure Use If rods failed to fully insert and RPS did not deenergize, Obj V. an operator is directed to pull RPS fuses. However, this See AOI-1 00-2 is beyond the actual design base Attachment 11 Transfer reactor pressure control to Panel 25-32 to allow Note: System Status for pressure control while the rest of the panel checklist prior to abandonment is being complete maintained by GOI-300-1 checklist . Before any transfer switch is placed in EMERGENCY, its Obj. V. associated control switch must be verified to be in the proper Obj. V. position. Placing a transfer switch in the EMERGENCY position enables the local control switch, and the device will assume the condition called for by the local control switc For example, if a transfer switch for an ADS valve is placed in EMERGENCY with the local control switch in OPEN, the ADS valve will ope Place the transfer switches for the ADS valves, and TP-1 the disconnect switches for the non-ADS valves in Obj. V. EMERGENCY after making sure the control switches are in the AUTO position. This action disables the Control Room hand switches and the ADS function and is performed to prevent spurious blowdown of the primary system. The other 3 SRVs are disabled by opening their breakers on 250VDC RMOV board 2B(3B).
Four ADS valves can be controlled from Panel Obj. V. . Six SRVs (Non-ADS) have only Obj. V. disconnect switches at Panel 25-3 OPl171.009 Revision 10 Appendix 0 Page 56 of 62 TP-4: ADS VALVE CONTROL CIRCUIT PCV 1-22
BFN Control Room Abandonment 2-AOI-100-2 Unit 2 Rev. 0051 Page 8 of 95 Unit 2 Subsequent Actions
[1] IF ALL control rods were NOT fully inserted AND RPS failed to deenergize, THEN (Otherwise N/A)
DIRECT an operator to Unit 2 Auxiliary Instrument Room to perform Attachment 1 D NOTES 1) The following transfers Reactor Pressure Control to Panel 2-25-32 to allow for pressure control while completing the Panel Checklis ) Attachment 9, Alarm Response Procedure Panel 2-25-32, provides for any alarms associated with this instructio CAUTION Failure to place control switch in desired position prior to transferring to emergency position may result in inadvertent actuation of the componen [NER/C] Operation from Panel 2-25-32 bypasses logic and interlocks normally associated with the components. [GE SIL 326,S1] ('
[2] At Panel 2-25-32, PLACE the following MSRV control switches in CLOSE/AUTO:
Switch N Description 2-HS-1-22C MAIN STM LINE B RELIEF VALVE D 2-HS-1-5C MAIN STM LINE A RELIEF VALVE D 2-HS-1-30C MAIN STM LINE C RELIEF VALVE D 2-HS-1-34C MAIN STM LINE C RELIEF VALVE D (
12. RO 218000G2.1.24 002IMEMlT2Gl///218000G2.1.24//BOTHIBANK 12117/2007 RMS Given the following plant conditions:
* Unit 2 is at rated powe * A loss of '2A' 250 Volt RMOV Board has occurre Which ONE of the following describes the effect on the Unit 2 Automatic Depressurization System (ADS) valves and ADS logic?
A. BOTH Div I & II ADS logics are still operabl ALL ADS valves will operate automaticall ALL ADS valves can be manually operate B. BOTH Div I & II ADS logic is NOT operabl NO ADS valves will operate automaticall Four (4) ADS valves can still be operated manuall C!" Div I ADS logic operable; Div II ADS logic is NOT operabl ALL ADS valves will still actuate automaticall ALL ADS valves can still be operated manuall D. Div I ADS logic operable, Div II ADS logic is NOT operabl ADS logic is ONLY capable of opening 4 ADS valves automaticall Four (4) ADS valves can still be operated manuall KIA Statement: 218000 ADS 2.1.24 - Conduct of Operations Ability to obtain and interpret station electrical and mechanical drawings KIA Justification: This question satisfies the KIA statement by requiring the candidate to recall and interpret the electrical logic drawing of the ADS system to determine the effect of a loss of power to that logi References: OPL 171.043 rev, 13 pg 16 f. through n. and TP-2 Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam Page: lof2 2/19/2008
REFERENCE PROVIDED: None Plausibility Analysis: ( In order to answer this question correctly the candidate must determine the foIlowing: 1. '2B' 250V RMOV Board supplies Div 1, '2A' 250V RMOV Board supplies Div 1/. This is the opposite of conventional logi . '2A' 250V RMOV Board supplies only relay power and only in Div 1/. 3. '2B' 250V RMOV Board supplies BOTH Div I and Div 1/ logi . Four ADS valves have an alternate power supply that is NOT '2B' 250V RMOV Boar . 'PCV-1-22 & 1-30' are fed from '2A' 250V RMOV Board but auto transfer to an alternate power suppl C - correct: The power supply for the LOGIC and the solenoid valves is 250VD V RMOV Bd 'A' supplies Power for relays in system 1/ of ADS Logic. A Loss of 250V RMOV Bd 'A' would prevent system 1/ actuation. AIl ADS valves would stiIl Actuate via Division I logi A - incorrect: Div 1/ logic is inoperabl B - incorrect: Only Div 1/ ADS logic is inoperable. AIl ADS valves wiIl stiIl operate out of the other division. This is plausible if the operator believes that '2A' vs '28' 250V RMOV Board fed both division logic D - incorrect: AIl ADS valves would function in automatic and manuaIl Duplicated and modified RO 218000G2.1.24 #1 due to it being used on the 0610 audit exam. Changed to a loss of '2A' 250V RMOV BD Page: 2of2 2/19/2008
OPL 171.043 Revision 13 Page 16 of 30 INSTRUCTOR NOTES The power supply for the LOGIC and the solenoid All ADS valves valves is 250VDC with alternate V RMOV Bd B supplies LOGIC Power for both power supplies system I & II can be manually A Loss of 250V RMOV Bd B would prevent actuation operated from backup control V RMOV Bd A supplies Power for relays in system panel (25-32) II of ADS Logic A Loss of 250V RMOV Bd A would prevent system II actuation PCV 1-22 is powered from 250V RMOV Board 2A with See section F. Unit alternate supply from 250V RMOV Board 2B Differences for U-3 Power Supplies PCVs 1-19, 1-31 are powered from 250V RMOV Board 2B. There is no alternate power to these valves PCVs 1-5 and 1-34 are normally powered from 250V RMOV Board 2C with alternate power supply from Battery Board 1 panel PCV 1-30 is normally powered from 250V RMOV Board 2A with a first alternate to 250V RMOV Board 2C and a second alternate to Battery Board 1 panel 7 Valves powered from 250V RMOV Bd 2C required alternate sources due to RMOV Board 2C not being environmentally qualified for a line break in secondary containment The transfer occurs automatically when undervoltage DCN 51106 relays (mounted on panel 2-25-32) sense a loss of power to 250V RMOV Bd 2 B. Instrumentation SRV discharge piping temperatures are measured by a multipoint recorder in the Control Room located on Panel 9-47 (range 0-600°F)
OPL 171.043 Revision 13 Page 15 of 30 INSTRUCTOR NOTES 4) EOls will direct the operator when this action is FLAGGING appropriate. Both keylocks must be placed in inhibit to prevent ADS blowdown 5) ADS Logic can be inhibited by removing fuses in Panel 9-30 in Auxiliary Instrument Roo ) The fuses for "A" logic are on terminal block "BB 104 & 105" (FU2-1-2EK3) 7) The "B" logic fuses are on terminal "AA94 & 95" (FU2-1-2EK13) 8) The time delay setting is chosen to be long 3-WAY enough so that HPCI has time to start and yet COMMUNICATIONS not so long that Core Spray and LPCI are unable to adequately cool the fuel if HPCI should fail to start The 100 psig and 185 psig ECCS interlocks are Obj. V.BA provided to ensure that there is a vessel level Obj. V. inventory medium available prior to initiating blowdown Ob V.CA of steam from the vessel Ob V. Ob V.EA ADS Trip Systems Ob V.BA Redundant trip systems from the same power supply Ob V. Ob V. AlC interlock ensures ADS functions when needed Ob V. There are two channels in each trip system 1) A and C in System I 2) Band Din System II Both channels of a trip system are required to function to initiate ADS from a given trip system This two-channel interlock is called the A-C interlock Ob V. and is provided to ensure that all signals to initiate Ob V. ADS response are confirmed, thus preventing an ADS Ob V response from an erroneous or failed signal Ob V. OPL 171.043 Revision 13 Appendix C Page 28 of 30 2ES1 E - CONTROL SWITCH PANEL 9-3 2E 1E 2EK9 - SYSTEM I, LOGIC C 3EK6 - SYSTEM I, LOGIC A 2EK17 - SYSTEM II, LOGIC B PANEL 9-30 2EK20 - SYSTEM II, LOGIC D (AUXINST ROOM) 2EK6 2EK9 2ES6-CONTROL SWITCH 2ES5E - TRANSFER RLY-1-XX RL Y-1-22 SWITCH I I t--------.
(xx) SRV Number 2E-K32 -r. _. . *--_*_*_*_*_* -*_*-*-*-i 1st Al III---..,......; i !i * , i "' 2ES5E 250vDC lNormal ! Normal MOV Bd. 2B !Emergency i 2ES6E : I !
I I PANEL 25-32
; 2ES5-E !
!i i 250vDC 2E-K32
. i RMOV i Normal 2ES5E B A Shown energized i Emergency Normal
}J-------'
_ - - L . . _ y t -_ _---' ____________________________ .. ______________________________ . ___________________ .. _______ _ Normal 2E-K32 SOLENOID ENERGIZED TO OPEN VALVE LOCAL Energizes Relay 2E-K32 which causes contacts to change position* 2- 730E929- 3 TP-2 ADS Valve Control Circuit PCV-1-22
13. RO 223002A2.06 OOllC/A/T2GIIPCISI!223002A3.01//RO/SROIBAN Given the following plant conditions:
* During performance of 2-SR-3.3.1.1.13, "Reactor Protection and Primary Containment Isolation Systems Low Reactor Water Level Instrument Channel B2 Calibration," 2-LlS-3-203D fails to actuat * It is determined that the failure is due to an inoperable switch and a replacement is NOT available for 4 day * The Shift Manager has determined that the proper action is to trip the inoperable channel ONL Which ONE of the following describes how this is accomplished and the effect on Unit status?
A. 01 Remove the fuse associated with 2-LlS-3-203D. A half scram will result and NO Primary Containment Isolation Valves will realig Remove the fuse associated with 2-LlS-3-203D. A half scram will result and PCIS Groups 2, 3 and 6 Inboard Isolation Valves will clos Place a trip into the Analog Trip Unit associated with 2-LlS-3-203D. NO half scram will result and NO Primary Containment Isolation Valves will realig Place a trip into the Analog Trip Unit associated with 2-LlS-3-203D. A half scram will result and PCIS Groups 2, 3 and 6 Outboard Isolation Valves will close.
Page: 1 of 3 2119/2008
KIA Statement: 223002 PCIS/Nuclear Steam Supply Shutoff A2.06 - Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abn cond or ops. Containment instrumentation failures KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the effect of an instrumentation failure and the corrective actions required as a result of that failur References: 2-01-99 Illustration 3 (page 6 of 11) 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 outcom NRC Exam Page: 20f3 2/19/2008
REFERENCE PROVIDED: None Plausibility Analysis: ( In order to answer this question correctly the candidate must determine the following: 1. Recognize the appropriate action is to ensure trip input by de-energizing the level switche . Recognize the effect on RPS logic based on the trip inpu . Recognize the effect on PCIS logic based on the trip inpu A - correct: Pulling the fuse associated with L1S-:-3-203D deenergizes th 'B2' Channel of RPS causing a 1/2 scram in RPS 'B' and a 1/4 isolation in PCIS. No PCIS valves will repositio B - incorrect: PCIS logic causes a "1/4-isolation" signal BUT no PC IV devices actuate. This is plausible because the action to ensure the trip input is correct. In addition, the "1/4-isolation" applies to the PCIS groups identified in the distracto C - incorrect: The method of inputting the trip is incorrect. Tripping an ATU cannot be ensured via a clearance. This is plausible because the RPS and PCIS response is correc D - incorrect: The method of inputting the trip is incorrect. Tripping an ATU cannot be ensured via a clearance. This is plausible because the "1/4-isolation" applies to the PCIS groups identified in the distractor. This distractor is also similar to answer "B;" except it is applied to outboard PCIV Amplification: Basically, when a signal received from a sensor reaches the ATU, it is compared to the setpoint value programmed into the ATU. When the setpoint is exceeded, the ATU transmits a TRIP output to the associated logic, whether RPS or PCIS. Essentially converting an ANALOG input into a DIGITAL output. The ATU has a feature which allows it to generate a TRIP output for testing purposes. It could serve the same function as removing power from the input sensor, but is not considered reliable enough to comply with Tech Spec requirements for "placing a channel in the TRIP condition."
Page: 3 of3 2119/2008
~.
BFN Reactor Protection System 2-01-99 Unit 2 Rev. 0073 PaQe 72 of 77 Illustration 3 (Page 6 of 11) Actions to Place RPS Instruments in Tripped Conditions (TS Table 3.3.1.1-1) DEVICE FUSE RELAY PANEL PRINT ALARMS REMARKS 2-LlS-3-203A 2-FU1-3-203AA 2-RL Y-099-05AK06A 9-15 2-730E915-9 ALARMS AND 1/2 SCRAM IN CHANNEL A 2-XA-55-4A-2 RXWATER (5AF6A) 2-RLY-099-5A-K25A 2-730E927-7 NO PCIS DEVICES ACTUAT RLY-OB4-1BAK5A RX VESSEL WTR LEVEL LOW LEVEL LOW 2-45EB71-2B HALF SCRAM (Level 3) 2-RLYcOB4-1BAK6A 2-XA-55-5B-1 A1 CHANNEL REACTOR CHANNEL A AUTO 1 channel actuated for secondary SCRAM Function: 4 containment and CREV initiation 2-LlS-3-203B 2-FU1-3-203BA 2-RL Y-099-05AKOBB 9-17 2-730E915-10 ALARMS AND 1/2 SCRAM IN CHANNEL B 2-XA-55-4A-2 RXWATER (5AF6B) 2-RLY-099-5A-K25B 2-730E927-8 NO PCISDEVICES ACTUAT RX VESSEL WTR LEVEL LOW LEVEL LOW 2-RLY-064-16AK5B 2-45E671-38 HALF SCRAM (Level 3) 2-RLY-064-16AK6B 2-XA-55-5B2 B1 CHANNEL REACTOR CHANNEL B AUTO 1 channel actuated for secondary SCRAM Function: 4 containment and CREV initiation 2-LlS-3-203C 2-FU 1-3-203CA 2-RLY-099-05AK06C 9-15 2-730E915-9 ALARMS AND 1/2 SCRAM IN CHANNEL A 2-XA-55-4A-2 RXWATER (5AF6C) 2-RL Y-099-5A-K25C 2-730E927 -7 NO PCIS DEVICES ACTUAT RX VESSEL WTR LEVEL LOW LEVEL LOW 2-RLY-OB4-1BAK5C 2 -45 E671-32 HALF SCRAM (Level 3) 2-RLY-064-1BAKBC 2 -XA-55-5 B-1 A2 CHANNEL REACTOR CHANNEL A AUTO 1 channel actuated for secondary SCRAM Function: 4 containment and CREV initiation 2-LlS-3-203D 2-FU1-3-203DA 2-RL Y-099-05AKOBD 9"17 2-730E915-10 ALARMS AND 1/2 SCRAM IN CHANNEL B 2-XA-55-4A-2 RXWATER (5AFBD) 2-RL Y-099-5A-K25D 2-730E927-8 NO PCIS DEVICES ACTUAT RX VESSEL WTR LEVEL LOW LEVEL LOW 2-RLY-064-16AK5D 2-45E671-44 HALF SCRAM (Level 3) 2-RLY-OB4-1BAKBD 2-XA-55-5B2 B2 CHANNEL REACTOR CHANNEL B AUTO 1 channel actuated for secondary SCRAM Function: 4 containment and CREV initiation NOTE: Device Function corresponds to the TS Table 3.3.1.1 Function BFN Primary Containment System 2-01-64 Unit 2 Rev. 0106 Page 102 of 194 Illustration 2 (Page 1 of 10) Actions to Place PC IS in Tripped Condition NOTE Water level designators (1-8)are listed for relationship to the applicable device onl (T.S. Tables 3.3.6.1-1,3.3.6.2-1, & 3.3.7.1-1) DEVICE FUSE RELAY PANEL PRINT ALARM REMARKS 2-LlS-3-203A 2-FU 1-3-203AA 5AK6A 9-15 2-730E915-9 2-XA-55-4A-2 ALARMS AND 1/2 SCRAM IN RXWATER (5A-F6A) 5AK25A 2-730E927-7 RX VESSEL WTR LEVEL LOW HALF SCRAM CHANNELA. CAUSES1~ LEVEL LOW 16AK5A 2-45E671-26 2-XA-55-5B-1 ISOLATION IN PC IS GROUPS 2,3, (Level 3) REACTOR CHANNEL A AUTO SCRAM 6 AND 8. NO PCIS DEVICES 16A6A ACTUATE.
2-LlS-3-203B 2-FU 1-3-203BA 5AK6B 9-17 2-730E915-10 2-XA-55-4A-2 ALARMS AND 1/2 SCRAM IN RXWATER (5A-F6B) 5AK25B 2-730E927-8 RX VESSEL WTR LEVEL LOW HALF SCRAM CHANNEL B. CAUSES 1/4 LEVEL LOW 2-45E671-38 2-XA-55-5B-2 ISOLATION IN PCIS GROUPS 2,3, 16AK5B (Level 3) REACTOR CHANNEL B AUTO SCRAM 6 AND 8. NO PCIS DEVICES 16AK6B ACTUATE.
2-LlS-3-203C 2-FU 1-3-203CA 5AK6C 9-15 2-730E915-9 2-XA-55-4A-2 ALARMS AND 1/2 SCRAM IN RXWATER (5A F6C) 5AK25C 2-730E927-7 RX VESSEL WTR LEVEL LOW HALF SCRAM . CHANNEL A. CAUSES 1/4 LEVEL LOW 2-XA-55-5 B-1 ISOLATION IN PC IS GROUPS 2,3, 16AK5C 2-45E671-32 (Level 3) REACTOR CHANNEL A AUTO SCRAM 6 AND 8. NO PCIS DEVICES 16AK6C ACTUATE.
2-LlS-3-203D 2-FU 1-3-203DA 5AK6D 9-17 2-730E915-10 2-XA-55-4A-2 ALARMS AND 1/2 SCRAM IN RXWATER (5A-F6D) 5AK25D 2-730E927-8 RX VESSEL WTR LEVEL LOW HALF SCRAM CHANNEL B. CAUSES 1/4 LEVEL LOW 2-45E671-44 2-XA-55-5B-2 ISOLATION IN PCIS GROUPS 2,3, 16AK5D (Level 3) REACTOR CHANNEL B AUTO SCRAM 6 AND 8. NO PCIS DEVICES 16AK6D ACTUATE.
Table 3.3.6.1-1: Function 2a and 5h Table 3.3.6.2-1: Function 1 Table 3.3.7.1-1: Function 1 ----- -- --
14. RO 223002A3.01 002/elAlT2G 11ADSIB5/223002A2.06//RO/SROIBANK Accident conditions on Unit 1 have resulted in an EOI-directed Emergency Depressurization. Reactor pressure is currently 106 psig. ALL systems function as designe Regarding the High Pressure Coolant Injection (HPCI) System, which ONE of the following statements describes the current status of the respective indications on the Containment Isolation Status System (CISS) Panel and Panel 9-3 (assume ALL instruments/controls actuate at the setpoint)? The 'amber' HPCI AUTO-ISOL LOGIC A/B lights, on Panel 9-3, are (1) The 'amber' (HPCI) PCIS LOGIC AlB INTITIATION lights, on the CISS Panel, are (2) The 'green' (HPCI) PCIS LOGIC A/B SUCCESS lights, on the CISS Panel, are (3)
(1 ) (2) (3) Extinguishe Extinguishe Extinguishe Illuminate Illuminate Illuminate Illuminate Extinguishe Illuminate D ..... Extinguishe Illuminate Illuminate KIA Statement:
223002 PCIS/Nuclear Steam Supply Shutoff A3.01 - Ability to monitor automatic operations of the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF including: System indicating lights and alarm KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to correctly determine the effect a loss of indication has on MSRVoperabilit References: OPL 171.042 rev, 19 pg 38 of 67 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 outcom NRC Exam Page: 10f2 2/19/2008
REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Unit differences between the setpoints for HPCllsolations between Unit 1 and Units 2/ . HPCI Isolation Logic differences in relation to the amber AUTO-ISOL LOGIC AlB lights, on Panel 9- . Differences between HPCI and RCIC Isolation logic, for Low Reactor Pressure Isolation, in relation to the amber AUTO-ISOL LOGIC AlB lights, on Panel 9- D - correct: This question hinges upon understanding NOT ONLY Unit differences; but, also logic differences between HPCI and RCIC, which are very similar system Their logic is often confused. First, the candidate must determine if the Low Reactor Pressure Isolation exists (This isolation actuates when dropping BELOW the setpoint).
The given stem conditions of "Unit 1" and "Reactor pressure is currently 106 psig" complicate the decision; as, the setpoint is above the procedural isolations for Units 2 and 3, but below the procedural setpoint for Unit Once the determination of a valid isolation signal is made, then the candidate must determine how this affects operator indications. The Nuclear Steam Supply Valves to the HPCI system will, in fact, automatically isolate (CLOSE). The nuance here is that although the isolation logic functions, the 'amber' HPCI AUTO-ISOL LOGIC AlB lights, at the HPCI control station on Panel 9-3, will NOT illuminate. This plays on a system difference, in that RCIC has the same exact lights ('amber' RCIC AUTO-ISOL LOGIC AlB) in the same relative location, which WILL illuminate on the Low Reactor Pressure Isolation. This is often a point of confusion to experienced operator The Containment Isolation Status System (CISS) Panel amber/green lights that correspond to (HPCI) PCIS LOGIC AlB INTITATION / SUCCESS will BOTH illuminate to provide the operator a visual cue that HPCI has BOTH a demand for isolation and that it has isolated successfull A - incorrect: If the candidate assumes that the Unit 1 HPCI Isolation is 105 psig (Unit 2/3) versus 110 psig, with the given stem addition of "(assume ALL instruments / controls actuate at the setpoint)," they will come to the conclusion that the isolation has NOT yet occurred. The result would be that this particular distractor will be chose B - incorrect: If the candidate confuses HPCI and RCIC logic light indications for the exact same scenario (Le., 'amber' RCIC AUTO-I SOL LOGIC AlB would be illuminated), then this particular distractor will be chose C - incorrect: If the candidate confuses the set of 'amber' (HPCI) PCIS LOGIC AlB INTITATION with the set of 'amber' HPCI AUTO-ISOL LOGIC AlB lights, then this particular distractor will be chosen.
Page: 2of2 2/19/2008
OPL 171.042 Revision 19 Page 38 of 67 INSTRUCTOR NOTES (2) Suction valve from CST (73-40) opens Mgmt Exp 06-003: if it was closed. (Provided at least one Once HPCI T&P'd torus suction valve is closed). iaw EOI App. 4, must evaluate per (3) HPCI pump discharge valves to CAUTION #5 prior feedwater system (73-34 and 73-44) to re-start. Read receive open signa .
(4) Test line isolation valves (73-35 and EFFECTIVE 73-36) close if they were ope COMMUNICATIO N (5) The steam supply valve to the turbine (73-16) open (6) The auxiliary oil pump and GSC blower are started by initiation signa (7) As oil pressure increases, the turbine control and stop valves open when hydraulic pressure is available to admit steam to the turbin (8) The minimum flow bypass valve (73-30) will be shut automatically when HPCI flow becomes adequat (9) The turbine control system will maintain turbine speed to provide constant flow.
2. HPCI Isolation TP-10 Obj. V.B. a. Conditions causing HPCI isolation Obj. V.C. Obj. V. (1 ) Low reactor pressure 105 psig (one Obj. V. out of two twice). 110 psig for Unit Only signal that DCN 51237 does not seal in Unit Difference (2) High HPCI area temperature ~170°F (Torus Area) or ~190°F (HPCI Pump SER 3-05 Room) (one out of two twice)
(3) High HPCI steam line flow (200%) (3-second time delay) (one out of two)
OPL 171.042 Revision 19 Appendix C Page 63 of 67 TP-9: HPCI Turbine Trip Logic
OPL 171.042 Revision 19 Appendix C Page 64 of 67 TP-10: HPCllsolation Logic
15. RO 239002A3.03 OOl/CIA/T2GIIMAIN STEAM/CIA/239002A3.03IIRO/SROIBANK Given the following plant conditions:
* The reactor is operating at 100% power and 1000 psi * A Turbine Control Valve malfunction resulted in Safety Relief Valve (SRV) '1-4' lifting and failing to resea Which ONE of the following describes the expected SRV tailpipe temperature?
REFERENCE PROVIDED KlA Statement: 239002 SRVs A3.03 - Ability to monitor automatic operations of the RELIEF/SAFETY VALVES including: Tail pipe temperatures KlA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the expected tailpipe temperature of an open MSRV using steam table References: Steam Tables 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 outcom NRC Exam Page: lof2 2/19/2008
REFERENCE PROVIDED: Steam Tables Plausibility Analysis: In order to answer this question correctly the candidate must: 1. Use the Steam Table Mollier Diagram to determine the correct process and temperature for an open MSR NOTE: This question is typical for a GFES examination, however the KIA provides little lattitude for a question with discriminatory value based on reading a multi-point recorder. In addition, with it's direct connection to an issue identified following the accident at TMI, the importance of understanding this process becomes self-eviden B - correct: This is a throttling process and is therefore isenthalpi A - incorrect: This temperature is indicative of saturation temperature for steam at tailpipe pressure (atmospheric).
C - incorrect: 345°F would be incorrectly determined if the candidate considered the process to be isenthalpic to the saturation line, then followed the constant superheat line to atmospheric pressur D - incorrect: This temperature is indicative of saturation temperature for reactor pressure. The operator may assume that due to the location of the tailpipe temperature detector that the indication would be consistent with the saturation temperature of the reactor coolant syste Amplification: Regarding Distractor "0", the reason it is plausible is because this is inline with what the operators at TMI expected to see and why they assummed the PORV was closed when it was still open.
Page: 2of2 2/19/2008
IN TI EFERE CE RO 10 TO ANDI TE (
Combustion Engineering Steam Tables 16. RO 239002A4.0S OOlIMEM/T2G1I32A-1I41239002A4.0SIIRO/SROIBANK Given the following plant conditions:
* Unit 2 is operating at 100% powe * A complete loss of Drywell Control Air occurs (BOTH headers).
- NEITHER crosstie with Containment Atmosphere Dilution (CAD), nor plant Control Air, can restore system pressur Which ONE of the following statements describes the effect on pneumatically operated valves inside the Primary Containment in accordance with 2-AOI-32A-1, "Loss of Drywell Control Air"? ALL MSRV's can still be cycled five times due to the air system check valve arrangemen B." ADS MSRVs can still be cycled five times due to the air system accumulator and check valve arrangemen Inboard MSIVs can be opened one time due to the air system accumulator and check valve arrangement to allow using the condenser as a heat sin Inboard MSIVs will remain OPEN due to spring pressure until closed by a valid
{ isolation signal due to the air system accumulator and check valve arrangemen c," KIA Statement: 239002 SRVs A4.08 - Ability to manually operate and/or monitor in the control room: Plant air system pressure: Plant-Specific KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the operability of MSRVs following a loss of pneumatic suppl References: 2-AOI-32A-1 Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam Page: lof2 211912008
REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Recognize that ADS MSRV accumulators allow for five cycle operation . Recognize that Inboard MSIVs are capable of being CLOSED one time, but not OPENED one time and that the MSIVs fail closed on loss of control ai B - correct: The ADS MSRV air accumulators are provided to ensure that the valves can be held open following failure of the air supply to the accumulators, and they are sized to contain sufficient air for a minimum of five valve operations. Operations of the ADS MSRV should be limited to 5 time A - incorrect: Only ADS MSRVs have accumulators sufficient to cycle five times. The remaining MSRVs will not function without a pneumatic suppl C - incorrect: The MSIVs fail closed on a loss of control air. This question is plausible if the candidate misunderstands the purpose of the MSIV accumulato D - incorrect: The MSIVs are designed to fail closed on a loss of control air. Spring pressure and air close the MSIV. Also a loss of contol air will cause the MSIV to drift close due to spring pressure. This distractor is plausible if the candidate thinks that spring pressure assists in opening the MSI LOK changed to MEMORY.
Page: 20f2 2/19/2008
BFN Loss of Drywell Control Air 2-AOI-32A-1 Unit 2 Rev. 0021 Page 5 of 9 OPERATOR ACTIONS Immediate Actions None Subsequent Actions
[1] IF ANY EOI entry condition is met, THEN ENTER the appropriate EOI(s). o NOTES 1) The MSIV air accumulators are designed to provide for one closing actuation following loss of air supply. Once closed the valve is held closed by the spring ) The ADS MSRV air accumulators are provided to assure that the valves can be held open following failure of the air supply to the accumulators, and they are sized to contain sufficient air for a minimum of five valve operations. Operations of the ADS MSRV should be limited to 5 time ) Nitrogen Tanks supply pressurized nitrogen to the Drywell Control Air System via the DWCA SUPPLY REGULATORS 2-PREG-32-49A and 2-PREG-32-49A (lead regulator will be set at 100 psig and backup regulator set at 5-8 psig lower)
4) DWCA NITROGEN REG STATION BYPASS VLV, 2-BYV-032-0141 can be used to maintain approximately 98 psig in DWCA Receiver Tanks A & B when required by plant conditions
[2] CHECK Drywell Control Air System operating properl REFER TO 2-01-32 o [3] IF Operation with DWCA Nitrogen Regulattion Bypass Valve OpenlThrottled is required, THEN REFER TO 2-01-32 o
1 RO 259002A4.03 002/C/AlT2G1IFW 0I-3//259002A4.03//ROINEW 02112/2008 Which ONE of the following describes the indications and controls available when the Unit Operator transfers the Unit 1 Reactor Fedwater Startup Level Controller from MANUAL to AUTO? ~ The Auto/Manual_---l(...:..1.L-)_ _ light extinguishes and the _ ......(= ) __ light illuminates .. The controller is now controlling in _ ......('-"'3.....
) _ _ Element with the controller output changed from (4) (1 ) (2) (3) (4)
A'I Amber; Blue; Single; Demand output to Level Setpoint outpu B. Amber; Blue; Three; Level Setpoint output to Demand outpu C. Blue; Amber; Single; Level Setpoint output to Demand outpu D. Blue; Amber; Three; Demand output to Level Setpoint outpu KIA Statement: 259002 Reactor Water Level Control A4.03 - Ability to manually operate and/or monitor in the control room: All individual component controllers when transferring from manual to automatic mode KIA Justification: This question satisfies the KIA statement by requiring the candidate to understand the process and indications for transferring FWLC from manual to aut References: 1-01-3 Illustration 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 outcom NRC Exam New question for RO 259002A4.03#1 to better meet the KIA Page: 10f2 2/19/2008
REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. The indications when in auto and the operation of the controlle A - correct: The AUTO/MANUAL pushbutton is a momentary backlit pushbutton with a split lens backlight. AUTO is blue and MANUAL is amber. Pushing this button toggles the controls from AUTO to MANUAL, and vice versa. In AUTO, the output is the LEVEL SETPOINT. In MANUAL, the Operator has manual control (demand) with RAISE/LOWER pushbuttons. Also, the Startup Level controller only controls in SINGLE ELEMENT when in AUTO (ref. Illustration 1, Sec. 3.0 description).
B - incorrect: The Startup Level controller only controls in SINGLE ELEMENT when in AUTO. This is plausible because the desired FWLC control method is THREE ELEMENT. Also, the controller will be in LEVEL SETPOINT outpu C - incorrect: The blue light will be lit and NOT the amber. This is plausible because of the differences between the units. Unit's 2 & 3 have different controllers. Also, the controller will be in LEVEL SETPOINT outpu D - incorrect: The blue light will be lit and NOT the amber. Same plausiblity as 'C'. The Startup Level controller only controls in SINGLE ELEMENT when in AUT Amplification: Changed question to better meet the KIA Page: 20f2 2/1912008
BFN Reactor Feedwater System 1-01-3 Unit 1 Rev. 0009 Page 179 of 210 Illustration 1 (Page 2 of 3) RFW Startup Level Controller Panel Display Station COMPONENTS (REFER TO FIGURE 1) RAMP ENABLE This is a maintained backlit amber pushbutton that will change the RAISE/LOWER pushbuttons from "One Shot" to "Ramp" control. This button is pushed for ramp control LOWER This is a momentary non-backlit pushbutto MANUAL/AUTO This is a momentary backlit pushbutton with a split lens backlight. AUTO is Blue and MANUAL is ambe Pushing this button toggles the controls from AUTO to MANUAL and vice versa. In AUTO control the Output is the Level Setpoint output, and in MANUAL the Operator manual control with RAISE/LOWER pushbutton RAISE This is a momentary non-backlit puhbutto MANUAL RAMP Ramps the RFW Startup Level Controller demand by MODE 2% per second by using RAISE/LOWER pushbutton MANUAL Changes the RFW Startup Level Controller demand by
"ONE-SHOT" MODE 0.1 % per push of the RAISE/LOWER pushbuttons AUTO RAMP MODE Ramps the RFW Startup Level Controller Setpoint value by 0.25 inches per secon AUTO-"ONE-SHOT" Changes the RFW Startup Level Controller Setpoint MODE value by 0.20 inches per push of the RAISE/LOWER puhbuttons. DESCRIPTION TheRFW SU LVL CONT, 1-UC-3-53, is used from 0 psig to approximately 350 psig RX Pressure. The Panel Display Station(PDS) varies the position of the RFW START-UP LCV, 1-LCV-003-0053, to control Feedwater flow to the reactor. When AUTO on the PDS is selected, then Reactor Water Level is controlled in the Automatic Single Element Mode with the level setpoint adjusted by the Unit Operator utilizing the RAISE/LOWER push-buttons on the PDS. When MANUALon the PDS is selected, then RFW START-UP LCV, 1-LCV-003-0053, position is changed manually by the Unit Operator using the RAISE/LOWER push-buttons. The RFW START-UP LCV, 1-LCV-003-0053 can also be positioned locally by handwhee . RO 261000K3.03 001lCIAlT2GlICONTIPRIlV.B.8/261000K3.031IRO/SROINEW 10/16107 Unit 2 has experienced a LOCA with the following plant conditions: * Drywell pressure is 51 psig and risin * Suppression Chamber (Torus) water level is 18.5 fee * An Emergency Depressurization has been conducte * The Torus is being vented through Standby Gas Treatment (SGT) 'A' trai * Standby Gas Trains 'B' and 'C' are INOPERABLE due to common mode failure IF Standby Gas Train 'A' were to suffer the same common mode failure under these conditions, which ONE of the following would be the next sequential, EOI-directed step to exhaust the primary containment atmosphere?
Vent the _~(-,-1.L..)_ _ in accordance with 2-EOI Appendix 13, "Emergency Venting Primary Containment," _ _ _ _ _ _ _ _ _-->.:(2::...<)'--_ _ _ _ _ _ _ _ __ REFERENCE PROVIDED (1 ) (2) A." Torus; via the Hardened Wetwell Vent Syste Torus; allowing the Primary Containment Vent Ducts to fai Drywell; via the Hardened Wetwell Vent Syste Drywell; allowing the Primary Containment Vent Ducts to fail.
Page: lof3 2119/2008
KIA Statement: 261000 SGTS K3.03 - Knowledge of the effect that a loss or malfunction of the STANDBY GAS TREATMENT SYSTEM will have on following: Primary containment pressure: Mark-I&II KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific knowledge of the relationship between SGT and the inerting proces References: OPL 171.016, Primary and Secondary Containment Systems, 2-EOI-2, Primary Containment Control, 2-EOI Appendix 13, Emergency Venting Primary Containment 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 outcom NRC Exam REFERENCE PROVIDED: 2-EOI-2, Primary Containment Control Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Drywell pressure is approaching the 55 psig pressure limi . Recognize which available vent path does NOT require SGT to be OPERABL . Recognize that venting the Suppression Chamber is preferred over the Drywell to facilitate scrubbing of radioactive fission product . Recognize which vent paths are available to the Torus and Drywel . Recognize the implications of Torus Water Level as it relates to the chosen vent pat A - correct: First, let it be recognized that in any case that you are venting the Drywell OR Torus, you are, in effect, venting the respective other. In this particular case, the impact of losing the last available Standby Gas Train removes the viability of utilizing Appendix 12, Primary Containment Venting. Additionally, based on given stem conditions of Drywell Pressure and Torus Water Level, the Pressure Suppression Pressure (PSP) curve (EOI-2, Curve 6) has been exceeded; and Torus pressure CANNOT be maintained in the "safe area of the curve." Also identified in stem conditions, an Emergency Depressurization has been conducted, providing the allowance to travel farther down the Primary Containment Pressure "PC/P" leg of EOI- Once current Torus Water Level is applied to the equation, the candidate would be Page: 2of3 2/1912008
OFFSITE RADIOACTIVITY RELEASE RATE (APPX 13)" versus the other optional path "VENT THE OW IRRESPECTIVE OF OFFSITE RADIOACTIVITY RELEASE RATE (APPX 13)." Without Appendix 13 available to the candidate, they must now ascertain which vent path is applicable. NO additional failures are provided in the stem conditions which would preclude the subsequent venting of the Torus; utilizing the Appendix 13 designated path. This designated path for the Torus is the Hardened Wetwell Vent System, as opposed to the Appendix 13 designated Drywell path, which is allowing the Primary Containment Vent Ducts to fai B - incorrect: Although venting the Torus would be the directed path per EOI-2 (by given stem conditions), the procedurally-designated path for the Torus is the Hardened Wetwell Vent System, as opposed to the designated Drywell path, which is allowing the Primary Containment Vent Ducts to fail. Destructive venting of the Suppression Chamber is physically possible; but, is NOT procedurally authorized. As detailed above, the candidate is ONLY directed, in EOI-2, to "VENT THE SUPPR CHMBR IRRESPECTIVE OF OFFSITE RADIOACTIVITY RELEASE RATE (APPX 13)." It does NOT specify a particular pat C - incorrect: Although venting the Drywell is a directed path in EOI-2, it is NOT yet required (by given stem conditions). Additionally, the procedurally-designated path for the Drywell is allowing the Primary Containment Vent Ducts to fail, as opposed to the designated path for the Torus, which is the Hardened Wetwell Vent System. If the Drywell path is chosen, the candidate's ONLY direction, once again, is. to "VENT THE OW IRRESPECTIVE OF OFFSITE RADIOACTIVITY RELEASE RATE (APPX 13)." It does NOT specify a particular pat D - incorrect: This is a fully-functional, procedurally-allowed vent path for the Drywell; directed by EOI-2 and detailed in Appendix 13. If the' candidate does NOT apply ALL of the given stem conditions, primarily Torus Water Level LESS THAN 20 feet, an error will be made. Complicating this decision for the candidate is another Decision Block, in EOI-2, inquiring as to "CAN THE SUPPR CHMBR BE VENTED?" Page: 3 of3 2/19/2008
#2 PUMP NPSH ANP VORTEX LIMITS 30 L "'"
e '"[Z0: 20
"-
INITIATE OW SPRAYS USING ONLY PUMPS NOT REQUIRED ill 15 TO ASSURE ADEQUATE CORE COOLING BY CONTINUOUS
"'f'"
u 10 INJ (APPX 17B) L
& :0 '"
> 11 11.512 13 14 15 16 17 18 19 20 SUPPR Pl LVl (FT) NO L EMERGENCY RPV DEPRESSURIZATION IS REQUIRED (EOI-1, RC/P-4: C1-2: C1-21: C5-1) BLOWERS L PC/P-11 L MAINTAIN SUPPR CHMBR PRESS BELOW 55 PSIG L PC/p*12 =:JL NOT REQU!RED CONTINUOUS L NO L > { VENT THE SUPPR OHMBR IRRESPECTIVE OF OFF SITE RADIOACTIVITY RELEASE RATE (APPX 13)
\ L PC/P*15 NO L IS REQUIRED VENT THE DW IRRESPECTIVE OF OFFSITE I) RADIOACTIVITY RELEASE RATE (APPX 13)
L L PC/p*17 NO L
2-EOI APPENDIX~13 Rev. 5 Page 1 of 6
.------------------------.
2-EOI APPENDIX-13 EMERGENCY VENTING PRIMARY CONTAINMENT LOCATION: Unit 2 Control Room ATTACHMENTS: 1.Tools and Equipment 2.Vent System Overview 3.Hardened Vent Flow Path NOTIFY SHIFT MNGR./SED of the following:
* Emergency Venting of Primary Containment is in progres * Off-Gas Release Rate Limits will be exceeded. VENT the Suppression Chamber as follows (Panel 9-3): IF . . . . . . . EITHER of the following exists: * Suppression Pool water level CANNOT be determined to be below 20 ft, OR * Suppression Chamber CANNOT be vented, THEN ..... CONTINUE in this procedure at Step PLACE keylock switch 2-HS-64-222B, HARDENED SUPPR CHBR VENT OUTBD PERMISSIVE, in PER CHECK blue indicating light above 2-HS-64-222B, HARDENED SUPPR CHBR VENT OUTBD PERMISSIVE, illuminate OPEN 2-FCV-64-222, HARDENED SUPPR CHBR VENT OUTBD ISOL VL PLACE keylock switch 2-HS-64-22IB, HARDENED SUPPR CHBR VENT INBD PERMISSIVE, in PER CHECK ~lue indicating light above 2-HS-64-22IB, HARDENED SUPPR CHBR VENT INBD PERMISSIVE, illuminate OPEN 2-FCV-64-221, HARDENED SUPPR CHBR VENT INBD ISOL VL Eor APPENDIX-13 Rev. 5 Page 2 of 6 (continued from previous page) CHECK Drywell and Suppression Chamber Pressure lowerin MAINTAIN Primary Containment Pressure below 55 psig using 2-FCV-64-222, HARDENED SUPR CHBR VENT OUTBD ISOL VLV, as directed by SRO.
3. IF ..... Suppression Chamber vent path is NOT available, THEN ... VENT the Drywell as follows: NOTIFY SHIFT MNGR./SED that Secondary Containment integrity failure is possibl NOTIFY RADCON that Reactor Building is being evacuated due to imminent failure of Primary Containment vent duct EVACUATE ALL Reactor Buildings using P.A. Syste START ALL available SGTS train VERIFY CLOSED 2-FCV~64-36, DW/SUPPR CHBR VENT TO SGT (Panel 9-3). VERIFY OPEN the following dampers (Panel 9-25)
* 2-FCO-64-40, REACTOR ZONE EXH TO SGTS * 2-FCO-64-41, REACTOR ZONE EXH TO SGT VERIFY CLOSED 2-FCV-64-29, DRYWELL VENT INBD ISOL VALVE (Panel 9-3 or Panel 9-54). DISPATCH personnel to Unit 2 Auxiliary Instrument Room to perform the following:
1) REFER TO Attachment 1 and OBTAIN one 12-i Banana Jack Jumper from Eor Equipment Storage Bo ) LOCATE terminal strip DD in Panel 9-43, Fron ) JUMPER DD-76 to DD-77 (Panel 9-43).
4) NOTIFY Unit Operator that jumper for 2-FCV-64-30, DRYWELL VENT OUTBD ISOLATION VLV, is in plac VERIFY OPEN 2-FCV-64-30, DRYWELL VENT OUTBD SOLATION VLV (Panel 9-3).
~, TO REACTOR BLDG VENT SYSTEM OVERVIEW EXHAUST FANS
~
§9 I ~I<iII I .. ~ ... - ~TOSGTS
<D
'".. <D 64-30 II 64-29
~ ~ I ~t=0 REl
!I-§lt ~ _____ , l REl
<D I a 2"
! I DRYWELL 2" _ : I
: I
":l : I
! I
!:
I I ~
:I.j>.
--
C(l (Xl
~
I DRYWELL PURGE FANS AND FILTERS !r-- I I
[ru (J)
f"'
:I I
I ":l ____:
: ~ I I
,---------
rn ---~ I t~- ___ I I I .j>. _II
~ ~!4
4'---'-_ L - -...LI 18" ----, 64-33 64-32 >"d~i'0
>-3PJ(]) I REACTOR BLDG >-3 LQ <: t:r:I VENTILATION >' (]). 0 EXHAUST (1 H
::r:U1U1
:5: >'
t:r:I 0 '"d ZI-h '"d
>-3 t:r:I O'l Z i'0 t:J H
:x:
I I--' W
2-EOI APPENDIX-13 Rev. 5 Page 6 of 6 ATTACHMENT 3
~I~
II,... I gIl:! 9~~ J: .... 1~ I~ 1~ ~ I ~ ...J u.. .... Z [R] ~ W ~ > Cl ~ W Z ~ w Cl c::
~ S
~
~ ~
N
~I N~ ~ ~I I 0'1 ~ ~ ~ ~ ~
6~-W
.re ~
LAST PAGE
19. RO 262001K4.04 OOllCIA/SYS/ACDIST/3/262001K4.041IRO/SROIBANK Given the following plant electrical distribution alignment:
* 4kV Shutdown Bus 1'43' Switch is in MANUA * ALL 4kV Shutdown Board '43' Switches are in AUT * A fault on '1 A' 4kV Unit Board de-energizes 4kV Shutdown Bus Which ONE of the following represents how and when the '1A' 4kV Shutdown Board alternate supply breaker will auto close?
A (1 } transfer occurs when (2} voltage decays to less than 30%.
) (2) fast; 4kV Shutdown Bus 1 B ..... slow; 4kV Shutdown Bus 1 fast; '1 A' 4kV Shutdown Board slow; '1 A' 4kV Shutdown Board KIA Statement:
262001 AC Electrical Distribution K4.04 - Knowledge of A.C. ELECTRICAL DISTRIBUTION design feature(s) and/or interlocks which provide for the following: Protective relaying KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions and times to correctly determine the response of the AC distribution system to a fault which initiates protective relayin References: OPL171.036 rev11, pg, 31, 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 outcom NRC Exam Page: lof2 2/1912008
REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Whether the transfer is a fast transfer or slow transfe . Whether the low voltage is sensed on the line side of the breaker or the load side of the breake NOTE: The plausibility of the distractors is based on determining the answers to the above question B - correct: With Shutdown Bus transfer scheme in manual the Shutdown Bus will not fast transfer when a fault occurs on 1A 4kV Unit Board. This will cause a loss of voltage to the line side of the normal feeder to the 'A' 4kV Shudown Board. Therefore a slow transfer will occur on the 'A' 4kV Shudown Board when voltage decays to 30%. A - incorrect: Fast Transfers are MANUAL only. This is plausible because the undervoltage sensing location is correc C - incorrect: Fast Transfers are MANUAL only. The undervoltage is sensed on the Shutdown BUS side of the breake D - incorrect: The undervoltage is sensed on the Shutdown BUS side of the breake This is plausible because the transfer scheme is correc Note - removed the'S' from the '43S' switch. Any '43' switch is an auto/manual transfer switch. The'S' or 'Sx' designation is not required to answer this question correctly.
Page: 20f2 2/19/2008
OPL 171.036 Revision 11 Page 31 of 58 7. Shutdown Board Transfer Scheme The only automatic transfer of power on a Obj. V.B. shutdown board is a delayed (slow) transfe Obj. V.C. In order for the transfer to take place, the bus Obj. V.D. transfer control switch (43Sx) must be in Procedural AUTOMATI Adherence when transferring boards (1) Undervoltage is sensed on the line side of the normal feeder breake (2) Voltage is available on the line side of the alternate feeder breake (3) The normal feeder breaker then receives a trip signa (4) A 52b contact on the normal supply breaker shuts in the close circuit of the alternate feeder breaker, indicating that the normal breaker is ope (5) A residual voltage relay shuts in the close circuit of the alternate supply breaker, indicating that board voltage has decayed to less than 30 percent of norma (6) The alternate supply breaker then close The shutdown board transfer scheme is NORMAL seeking. If power is restored to the line side of the normal feeder breaker, and if the 43Sx switch is still in AUTOMATIC, then a "slow" transfer back to the normal supply will occu This will cause momentary power loss to loads on the bus and ESF actuations are possibl **b Manual High Speed (Fast Transfer) Obj. V.B. Obj. V.C. To fast transfer a shutdown board perform the Review INPO following: SOER 83-06
20. RO 262002Al.02 OOllCIA/T2/G llUNIT PREFFERRED/CIA 2.5/2.91262002AAl.02IBF05301IRO/SROIBANK Given the following plant conditions:
* Unit 3 is in a normal lineu * The following alarm is received: - UNIT PFD SUPPLY ABNORMAL * It is determined that the alarm is due to the Unit 3 Unit Preferred AC Generator Overvoltage condition Which ONE of the following describes the result of this condition?
Unit 3 Breaker 1001 _ _. l.-(1!. J.)_ _ Unit 2 Breaker 1003 _ _---=(2:.,L.) _ _,; and the Motor-Motor-Generator (MMG) set _ _ _ _ _ _ _. >,.;(3"-')_ _ _ _ _ _ __
(1) (2) (3)
A. trips OPEN; is interlocked OPEN; automatically shuts dow B. is interlocked OPEN; trips OPEN; automatically shuts dow C!' trips OPEN; is interlocked OPEN; continues to run without excitatio D. is interlocked OPEN; trips OPEN; continues to run without excitatio KIA Statement: 262002 UPS (AC/DC) A 1.02 Ability to predict andlor monitor changes in parameters associated with operating the UNINTERRUPTABLE POWER SUPPLY (A.C'!D.C.) controls including: Motor generator output KIA Justification: This question satisfies the KIA statement by requiring the candidate to correctly apply a specific operating condition of the UPS MMG Set to the correct response of the system to that conditio References: OPL 171.102, Rev.6, pg 20, 21, TP-2 and 3-ARP-9-8B, Rev.9, tile 35 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 solve a problem. This requires mentally using this knowledge and its meaning to resolve the proble NRC Exam Page: lof2 2/19/2008
REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: . 1. The 1001 and 1003 breakers from an MMG set will trip on overvoltage or underfrequency at the output of the MM . Unit 2 MMG Breakers are interlocked to prevent alternate power to unit 1 and 3 at the same tim . When an overvoltage condition exists at the Generator Output, the 1001 breaker from the MMG Set trip . Excitation is lost and the MMG Set continues to ru . The "Hold to build up voltage" switch must be depressed to restore voltag C - correct: The 1001 and 1003 breakers from an MMG set will trip on overvoltage or underfrequency at the output of the MMG. Also Unit 2 MMG Breakers are interlocked to prevent alternate power to unit 1 and 3 at the same tim A - incorrect: The MMG set does not automatically shut down. This is plausible because the breaker lineup is correc B - incorrect: The MMG set does not automatically shut down. This is plausible although the breaker lineup is backward D - incorrect: The breaker lineup is backwards. This is plausible because the MMG Set will continue to run without excitatio Note - The operator could incorrectly assume that the Unit 3 1001 bkr is interlocked to prevent paralleling battery boards 2 & 3 (ref TP-2).
Page: 20f2 2/19/2008
OPL171.102 Revision 6 Page 20 of69 (d) Another Unit's MMG set The second alternate is from another unit's MMG set output. Unit 2 MMG is the second alternate for either Unit 1 or Unit 3; Unit 3 is the second alternate for Unit Transfers to this source are done manually at Battery Board 2 panel 11.
b. MMG Sets (Unit 2&3) Obj. V.B. TP-11 (1) The MMG is normally driven By the Obj.v.D. AC motor, powered from 480V Obj.v.D.2.d/j Shutdown Board A. Should this Obj V.E. supply fail, the AC motor is Obj.v.E.2.d/i automatically disconnected and the Obj V.B. DC motor starts, powered from Obj.v.C. V Battery Board. The DC Obj.V.D. motor has an alternate power Obj.V.E. supply from another 250V Battery Board. Transfer to the alternate DC source is manua Underfrequency on the generator output will trip the DC moto Transfer of the MMG set back to the AC motor is manua (2) The 1001 and 1003 breakers from an MMG set will trip on overvoltage or underfrequency at the output of the G. Also Unit 2 MMG
OPL171.102 Revision 6 Page 21 of 69 (3) When an under frequency or Obj. V.B. overvoltage condition exists at the Obj. V.C. Generator Output the following Obj. V.D. occurs Obj. V.E. (a) BB panel 10 breakers from the MMG Set tri U2 1001 (U2) 1003 (U1 &3) U3 1001 (U3) 1003 (U2)
(b) Excitation is lost and the MMG Set continues to ru (The Hold to build up voltage switch must be depressed to restore voltage.)
OPL171.102 Revision 6 Page 58 of 69 OPL171.102 480V RMOV BD1A Norm Norm Norm 250V DC 250V DC Batt Bd 5 Batt Bd 4 All ) Batt Bd 6 IBr 4L.f---~LI-----' All
~ .I All 480V Bd 250V de 250V de SD Bd Batt Bd 5 I Rectifier I Batt Bd 6 3A l
I)
~~ ~~
I Invelrter I
)
r--------- --- --, Panel 101 ) I I )) I Panel 111
~nF~lnn-F~-r~~ I I Batt Bd 1 I L ________________ ~
480V RMOVBd 2B
~_~ ~(
To P ) Ir -- w02 -----1001- 1003'1
) 480V aoel1°~--I-nnn}n-j--~ RMId38 Paoel11 L--- ~~ ~--'--- ----- -f --O~3---- 1'--- ---1 Panel 10 I ) ~~02 001 I Panel 11 I
~--L------I~---C--~,
I I L ________________ I Batt Bd 3 ~ TP-2: POWER SUPPLIES TO UPS BATTERY BOARD CABINETS
8FN Panel 1-9-8 1-ARP-9-88 Unit 1 1-XA-55-88 Rev. 0009 Page 42 of 42 Sensor/Trip Point: UNIT PFD SUPPLY Relay SE - loss of normal DC power sourc ABNORMAL Relay TS - DCXfer switch transfers to Emergency DC Power Sourc Regulating Transformer Common Alar f35 1-INV-252-001 ,INVT~1 System Common Alarm.
~--------------~~
(Page 1 of 1)
Sensor EL 593' 250V DC Battery Board 2 Location: Probable Loss of normal DC power source Cause: DC power transfe Relay failure INVT-1 System Common Alarms 1. Fan Failure Rectifier 2. Over temperature Rectifier 3. AC Power Failure to Rectifier 4. Low DC Voltage 5. High DC Voltage 6. Low DC Disconnect 7 . Fan Failure Inverter 8. Alternate Source Failure 9. :Low AC Output Voltage 10. High Output Voltage 11. Inverter Fuse Blown 12. Static Switch Fuse Blown 13. Over Temperature Inverter E. PFD Regulating XFMR Common Alarms 1. Transformer Over temperature 2. Fan Failure 3. CB1 Breaker Trip 4. CB2 Breaker Trip Automatic A. Auto transfer to DC Power Source on Rectifier failur Action: B. Auto transfer to Alternate AC supply (Regulated Transformer) on Inverter failur Operator A. IF 120V AC Unit Preferred is lost, THEN Action: REFER TO 1-AOI-57-4, Loss of Unit Preferre o B. REFER TO appropriate portion of 0-OI-57C, 208V/120V AC Electrical Syste o References: 0-45E641-2 1-45E620-11 1-3300D 15A4585-1 10-100467 0-20-100756 20-110437
21. RO 263000Kl.02 OOllMEMlT2G1I250VDC/3/263000Kl.02//RO/SROIBANK Which ONE of the following statements describes the operation of '2B' 250VDC Battery Charger? The normal power supply to '2B' 250VDC Battery Charger is '2A' 480V Shutdown Boar '2B' 250VDC Battery Charger can supply, directly from unit 2 Battery Board Room, any of the six Unit & Plant 250VDC Battery Board Load shedding of the '2B' 250VDC Battery Charger CANNOT be bypassed until the Load Shed signal has been rese D." Load shedding of the '2B' 250VDC Battery Charger can be bypassed by placing the "Emergency ON Select Switch" in the "EMERGENCY ON" Positio KIA Statement: 263000 DC Electrical Distribution K1.02 - Knowledge of the physical connections and/or cause- effect relationships between D.C. ELECTRICAL DISTRIBUTION and the following: Battery charger and battery KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific knowledge of battery charger operatio References: OPL 171.037 rev 10, pg 11, 12 & 31 and 0-01-57 Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam Page: 10f2 2/19/2008
REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Normal and Alternate power to '2B' 250VDC Battery Charge . Loads capable of being supplied by '2B' 250VDC Battery Charge . Load Shedding Logic and bypass capabilit D - correct: 250V DC battery chargers 1, 2A and 2B wi" load shed upon receipt of a Unit 1 or Unit 2 accident signal and any Unit 1/2 shutdown board being supplied by its respective diesel generator or cross tied to a Unit 3 shutdown board and a unit three Diesel Generator. The load shedding feature can be bypassed by placing the
"Emergency" switch on the charger to the "EM ERG" position A - incorrect: This is plausible because '2A' 480V Shutdown Board is the Normal supply to '2A' 250VDC Battery Charge B - incorrect: This is plausible because '2B' 250VDC Battery Charger is capable of supplying any of the six 250V Battery Boards, but NOT directly from Unit 2 Battery Board Roo C - incorrect: This is plausible because the '2B' 250V Battery Charger is considered a spare charger and the candidate may draw the conclusion that the spare charger doesen't have the ablility to defeat the load shed logi Amplification: The actual switch position in '0' is "Emergency ON" in procedure 0-01-570. The position in the lesson plan is NOT correc Changed distractor 'C' for plausiblity.
Page: 2 of2 2/1912008
OPL 171.037 Revision 10 Page 11 of 70 (2) The Plant/Station Batteries (4, 5, and 6) are Obj V. Class Non-1 E and are utilized primarily for U-2, Obj. V. U-1, and U-3 respectively --for normal loads Obj. V. (3) Battery (4) Room is located on Unit 3 in the Turbine Building on Elev. 586 (4) Battery (5 & 6) Rooms are located on the Turbine Floor, Elev. 617 (5) The boards and chargers for the Unit Batteries are located in Battery Board Rooms adjacent to the batteries they serve, with the spare charger being in the Unit 2 Battery Board room. (Battery Boards 5 & 6 and their associated chargers are located adjacent to the batteries, but are in the open space of the turbine floor.) V Plant DC components (1) Battery charger (a) The battery chargers are of the solid state rectifier type. They normally supply loads on the 250V Plant DC Distribution System. Upon loss of power to the charger, the battery supplies the load (b) The main bank chargers only provide float and equalize charge when tied to their loads. The chargers are not placed on fast charge (high voltage equalize) with any loads attache (c) They can recharge a fully discharged battery in 12 hours while supplying normal load (d) Battery charger power supplies are Follow Procedure manual transfer only.
250V Batterv Alternate Source Normal Source Obj. V. Charqer (Charger Service bus) Obj. V. V SO Bd 1A 480V Common Bd 1 1 Obj .V. Comp6D Comp 3A 480V SO Bd 2A 480V Common Bd 1 2A Comp6D Comp 3A 480V SO Bd 2B 480V Common Bd 1 2B Comp6D Comp 3A 480V SO Bd 3A 480V Common Bd 1
Comp6D Comp 3A
OPL171.037 Revision 10 Page 12 of 70 480V SD Bd 3B 480V Common Bd 1
Comp6D Comp 3A 480V Com Bd 1 5 (no alternate) Comp5C 480V Com Bd 3 6 (no alternate) Comp3D 2B spare charger DC output can be directed to any of four feeders. Three DC outputs can be connected to battery board 1, 2, or 3. The fourth output is connected to a new output transfer TP-2 & TP-7 switch (located in battery board room 4) which charges batteries 4,5, or 6 plant batteries. A mechanical interlock permits closing only one output feeder at a time. (A slide bar is utilized in battery Attention to Detail board room 2 and a Kirk key interlock is used in battery board room 4)
OPL 171.037 Revision 10 Page 32 of 70 BB-3 supplies 250V RMOV Boards 3A, 1B, 2B.
The 250V DC Shutdown Board Battery System Five 120-cell batteries (one battery and battery charger for each Shutdown Board and one spare battery charger).One battery and charger for each Unit 1 and Unit 2 4KV Shutdown Board, one battery and charger for Unit 3 4KV Shutdown Board 3EB, and one spare portable charger.
The Units 1 and 2 Shutdown Board Batteries are located in the Reactor Building at Elev. 621, adjacent to the A and C 4KV Shutdown Board Rooms Batteries A and B are at the Unit 1 end (A Shutdown Board Room), Batteries C and 0 are at the Unit 2 end (C Shutdown Board Room) .. 3EB Shutdown Board Battery is located on the south end of Unit 3 Diesel Generator Building at Elev. 583.
The chargers are similar to those used in the 250V DC Unit and Plant System. Each charger has only one power supply. A, B, C, and 0 are powered from Units 1 or 2 480V Reactor MOV Boards; 3EB is supplied from 3EA 480V Diesel Auxiliary Board. All supply breakers are manual. (The portable charger has a plug-in type power lead for use with a 480V receptacle.)
480 SID BD Normal source Alternate source 1A SID B Batt. Dist. Panel SB-A Unit Batt. Bd. 2 1B SID B Batt. Dist. Panel SB-C Unit Batt. Bd. 3 2A SID B Batt. Dist. Panel SB-B Unit Batt. Bd. 1 2B SID B Batt. Dist. Panel SB-D Unit Batt. Bd. 3 3A SID B Unit Batt Bd 1 Unit Batt Bd 2 3B SID B Unit Batt Bd 3 Unit Batt Bd 1 480V Shutdown Boards 2B, 3A and 3B have remote transfer capabilities for 250V DC Control Power During an Appendix R Event. The transfer switch for 2B is located on 4KV SID Bd 0, and the switches for 3A & 3B are in Elv. 593 Elect. Board Room on Unit 3.
125VDC diesel battery system This DC system consists of eight batteries, sixteen chargers. Located in the room with the diesel generator. One of its chargers is also in the room; other is immediately outside.
1) Control and logic power 2) Governor booster pumps 3) Generator relay protection 4) Generator field flashing 5) Motor-driven fuel pump Review enabling objective OPL 171.037 Revision 10 Page 47 of 70
- -
-=
:::
.=. -- -
480" SO BO 1A NOR
............
BATTERY CHARGER N ALT ,.-....
............
480v SO BO 2A
.***.*...*..
BATTERY CHARGER No.2A ALT .............. 480" SO BO 2B
.............
BATTERY NOR * I CHARGER S U) No. 2B a:: UJ U) Z
************* 1
* rz l-I-
::J 480" SO BO 3A C I-NOR ._ ....... -... -- ::J
c:l N BATTERY 0 CHARGER I-N ............ 480" SO BO 3B NOR BATTERY CHARGER~----------~---4----~~----~--~--~--+----T--+4 N I-------l A LT BATT BATT BATT BATT BO 1 BO 2 B03 BD4 480" ....... _---_ .... __ ...... '-------- --- ----- '-- ------ -------- ... _- ................... __ .. .. COMMON BO 1 Tp . . 2 250V DC Power Distribution
BFN DC Electrical System 0-01-570 UnitO Rev. 0117 Page 16 of 247 PRECAUTIONS AND LIMITATIONS (continued) All safety requirements concerning smoking, fires or sparks should be observed when in the Battery-Battery Board Rooms because of potential accumulation of hydrogen in flammable amount V Unit Battery Charger 1,2A,2B and 3 Emergency ON select switch bypasses battery charger emergency load shed contacts. Placing the
, select switch in Emergency ON reestablishes charger operations with an accident signal present and Diesel Generator voltage available. Battery Charger 4 supply breaker, 480V Shutdown Board 3B, Compt 60, receives a trip signal from the load shed logic and the breaker must be manually re-closed after a 40 second time delay to restore the charger to service. The annunciation circuit for the 250V Unit Battery Charger 3 does NOT work when the EMER/OFF/ON Select Switch is in the EMER Positio [II/C] Neutron monitoring battery chargers are NOT stand alone power supplies and shall only be operated while connected to the neutron monitoring batterie [BFPER 940862] Within 30 minutes after the loss of the normal charger to a 250V Unit Battery another charger shall be placed in service to that battery and load reduced so that the battery is NOT dischargin [NRC/C] Upon return to service of 24V DC Neutron Monitoring Battery A or B, Instrument Maintenance must perform functional tests on SRMs and IRMs that are powered from the affected battery board (In that the IRMs and SRMs are normally inoperable after entering RUN mode due to lack of testing, these tests are N/A for the IRMs and the SRMs if the Unit is in RUN Mode and the IRMs and SRMs are inoperable). Prior to calling the IRMs and SRMs operable, the tests have to be performed. [NRC IE Inspect Follow-up Item 86-40] To return equipment to service following a failure or trip, the shutdown section of this instruction should be performed on the equipment failed. The initial conditions may NOT be applicable in this cas [NRC/C] The transfer of 250VDC control power to a 4kV Shutdown Board with a diesel generator operating may cause an inadvertent start of a RHRSW pum [LER 88021/25]
22. RO 264000K5.06 OOllCIAlT2GlI82 - DGI9/264000K5.06//RO/SROIBANK Given the following plant conditions:
* Unit 2 is operating at 100% Powe * No Equipment is Out of Servic * A large-break LOCA occurs in the drywell and the following conditions exist: - Drywell Pressure peaked at 28 psig and is currently at 20 psi Reactor Pressure is at 110 psi Reactor Water Level is at (-)120 inche Offsite power is availabl Which ONE of the following describes the associated equipment and proper loading sequence?
A." '28' RHR Pump and '28' Core Spray Pump start 7 seconds after the accident signal is receive . '2C' RHR Pump and '2C' Core Spray Pump start 7 seconds after the accident signal is receive '2A' RHR Pump starts immediately and '2A' Core Spray Pump starts 7 seconds after the accident signal is receive RHR Service Water Pumps, lined up for Emergency Equipment Cooling Water (EECW), start 14 seconds after the accident signal is receive KIA Statement: 264000 EDGs K5.06 - Knowledge of the operational implications of the following concepts as they apply to EMERGENCY GENERATORS (DIESEL/JET) : Load sequencing KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions and times to correctly determine the effect of load sequencing on plant equipment supplied by the Emergency Generator References: OPL 171.038 rev 16, pg 37 & 38 and 2-01-7 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 outcom NRC Exam Page: 1of2 2/19/2008
REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Load Sequencing is NVA (Normal Voltage Available) and NOT DGVA (DIG Voltage Available ). 2. Based on Item 1 above, the proper load sequencing with a Common Accident Signal (CAS) on Unit-2 alone and NOT in addition to a CAS on Unit A - correct: An accident signal with normal power available (NVA) causes the loads to sequence on in 7 second intervals (begining with the 'A' pumps) starting at time 0 after the accident signal is received. With NVA the EECW pumps start 28 seconds after the accident signal is reveive B - incorrect: This is plausible because 2-01-74 P&L 3.2.B defines the start time as 7 second "intervals".
C - incorrect: This is plausible because this is the sequence that they would start if a diesel generator was tied to a board (DGVA).
D - incorrect: This is plausible because RHRSW pumps all start at 14 seconds if load sequencing is DGV Amplification: Candidates ARE required to know the time, in seconds, for ECCS loading sequence under accident conditions.
Page: 20f2 2/19/2008
OPL171.038 Revision 16 Page 37 of 63 INSTRUCTOR NOTES (8) An OVERSPEED trip stops injector rocker arm motion, causing the engine to stop immediately. This will also cause a START FAILURE condition to be sensed by the control and alarm circuit (9) If the overspeed trip lever is reset, the engine will start again. This may be prevented by opening the LOGIC breaker and pushing both local engine stop push buttons, then waiting 15 minutes for all of the shutdown sequence timers to time out before resetting the trip leve (10) To reset the overspeed trip rotate the limit switch fully clockwise by hand and pull down on the reset lever. This should latch and keep the limit switch positioned clockwis (11) A local emergency stop may be Obj.V.D.17 required if the normal remote or local Obj.v.E.17 electrical stop signals fail to begin the engine stop sequence. This may be performed by pulling the fuel rack injector lever (located between the diesel and the engine control cabinet) outward from the engine, and holding until the engine comes to a stop. If the DG had been started using the local slow start procedure, a "start failure" signal may be sent to the redundant start signal, which would cause the engine to try and restart until the redundant start circuit locks out after 5.5 seconds elaps (12) The redundant start may be canceled before the start circuit locks out by opening the logic breaker and pushing both engine stop push-buttons. Note that pulling the engine driven fuel pump shutoff plunger will not stop the diesel since the electric fuel pump will still be supplying fuel. (01-82) 3. Accident Operation Accident signal received (CASx)
(1) Signals diesel generators to star OPL171.038 Revision 16 Page 38 of 63 INSTRUCTOR NOTES (2) Opens diesel output breakers if shu Obj.v. If normal voltage is available, load will Obj.v. Obj.v.D.15 sequence on as follows: (NVA)
Obj.v.E. 15 Time After Accident SID Board SID Board SID Board SID Board A C B D 0 RHR/CS A 7 RHR/CS B 14 RHR/CS C 21 RHR/CS D 28 RHRSW* RHRSW* RHRSW* RHRSW*
*RHRSW pumps assigned for EECW automatic start If normal voltage is NOT available: (DGVA) Obj.v. Obj.v. (1) After 5-second time delay, all 4kV Shutdown Board loads except 4160/480V transformer breakers are automatically trippe (2) Diesel generator output breaker closes when diesel is at spee (3) Loa d s sequence as .In d'Ica te d b eow I
Time After Accident SID Board SID Board SID Board SID Board A B C D 0 RHR A RHR C RHR B RHR D 7 CSA CS C CS B CS D 14 RHRSW* RHRSW* RHRSW* RHRSW*
*RHRSW pumps assigned for EECW automatic start Certain 480V loads are shed whenever an accident signal is received in conjunction with the diesel generator tied to the board. (see OPL 171.072)
BFN Residual Heat Removal System 2-01-74 Unit 2 Rev. 0133 Page 17 of 367 LPCI (continued) Upon an automatic LPCI initiation with normal power available, RHR Pump 2A starts immediately and 2B, 2C, 20 sequentially start at 7 second interval Otherwise, all RHR pumps start immediately once diesel power is available
'(and normal power unavailable). Manually stopping an RHR pump after LPCI initiation disables automatic restart of that pump until the initiation signal is reset. The affected RHR pump can still be started manually. Shutdown Cooling Prior to initiating Shutdown Cooling, RHR should be flushed to Radwaste until conductivity is less than 2.0 micromho/cm with less than 0.1 ppm chlorides (unless directed otherwise by 2-AOI-74-1, Loss of Shutdown Cooling). If CS&S has been aligned as the keep fill source for two days or more a chemistry sample should be requested and results analyzed to determine if flushing is require When in Shutdown Cooling, reactor temperature should be maintained greater than 72°F and only be controlled by throttling RHRSW flow. This is to assure adequate mixing of reactor wate . [NER/C] Reactor vessel water temperatures below 68°F exceed the temperature reactivity assumed in the criticality analysis. [INPO SER 90-017] [NER/C] Maintaining water temperature below 100°F minimizes the release of soluble activity. [GE SIL 541] Shutdown Cooling operation at saturated conditions (212°F) with 2 RHR pumps operating at or near combined maximum flow (20,000 gpm) could cause Jet Pump Cavitation. Indications of Jet Pump Cavitation are as follows: Rise in RHR System flow without a corresponding rise in Jet Pump flo . Fluctuation of Jet Pump flo . Louder "Rumbling" noise heard when vessel head is of Corrective action for any of these symptoms would be to reduce RHR flow until the symptom is correcte . RO 300000K2.02 OOlIMEMiT2GlICA//300000K2.02/2.8/2.8IRO/SROIBANK Which ONE of the following lists the current power supplies to the Control and Service Air Compressor motors? 'A' and 'B' are fed from 480V Common Board #1 'C' and '0' from 480V Shutdown Boards '1 B' & '2B', respectively 'G' from 4KV Shutdown Board 'B' and 480V Shutdown Board '2A' 'E' from 480V Common Board #1 'A' and '0' are fed from 480V Common Board #1 'B' and 'C' from 480V Shutdown Boards '1 B' & '2B', respectively 'G' from 4KV Shutdown Board 'B' and 480V RMOV Board '2A' 'F' from 480V Common Board #3 'A' is fed from 480V Shutdown Board '1 B' 'B' and 'F' from 480V Common Board #3 'C' from 480V Shutdown Board '1 A' '0' from 480V Shutdown Board '2A' 'G' from 4KV Common Board #2 0.'; 'A' is fed from 480V Shutdown Board '1 B' 'B' and 'C' from 480V Common Board #1 '0' from 480V Shutdown Board '2A' 'G' from 4KV Shutdown Board 'B' and 480V RMOV Board '2A' 'E' from 480V Common Board #3 KIA Statement:
300000 Instrument Air K2.02 - Knowledge of electrical power supplies to the following: Emergency air compressor KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific knowledge of the power supplies of ALL air compressor References: OPL 171.054 rev 12, pg 9, 10, 14 & 30 Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam Page: lof2 2/19/2008
REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Power supplies to the six air compressor NOTE: Regarding plausibility, all the power supplies listed in the distractors are capable of supplying power to an air compresso D - correct: The air compressor power supplies are as follows:
'A' is fed from 480V Shutdown Board '1 B' 'B' and 'C' from 480V Common Board #1 'D' from 480V Shutdown Board '2A' 'G' from 4KV Shutdown Board Band 480V RMOV Board '2A' 'E' from 480V Common Board #3 A - incorrect: B, G & E are correct. A, C & D are NOT correc B - incorrect: F & G are correct. A, B, C, & D are NOT correc C - incorrect: A, D & F are correct. B, C & G are NOT correct.
Page: 2of2 2/19/2008
OPL 171.054 Revision 12 Page 9 of 72 ( x. Lesson Body A. Control Air System 1. **The purpose of the Control Air System is to process ** SOER 88-1 and distribute oil-free control air, dried to a low dew point Obj. V. and free of foreign materials. This high-quality air is required throughout the plant and yard to ensure the proper functioning of pneumatically operated instruments, valves, and final operator . Basic Description of Flow Path TP-1 a. The station control air system has 5 air compressors, Obj. V. each designed for continuous operatio Obj. V. b. Common header (fed by air compressors A-O and G)
(1) The control air system is normally aligned with the The G air compressor G air compressor running and loaded. The will be discussed later in existing A.;.O air compressors are aligned with one this section of the lesson in second lead, one in third lead, and at least one pla compressor in standb (2) 3 control air receivers (3) 4 dual dryers One for each unit's control air header (units 1, 2 & 3 through their 4-inch headers) and One standby dryer supplies the standby, 3- inch common control air header for all normally aligned to all three units three units (4) Outlet from large service air receiver is connected to the control air receivers through a pressure control valve 0-FCV-33-1, which will automatically open to supply service air to the control air header if control air pressure falls to 85 psi c. 4-inch control air header (1 per unit) is supplied from TP-1 each unit dryer and backed up bya common, 3-inch standby heade . Control Air System Component Description a. Four Reciprocating Air Compressors A-O (2-stage, double acting, Y-type) are located EI 565, U-1 Turbine Buildin (1) Supply air to the control air receivers at 610 scfm each at a normal operating pressure of 90 - 101 psi (2) 480V, 60 Hz, 3-phase, drive motors (3) Power supplies A from 480V Shutdown Board 1B .
OPL171.054 Revision 12 Page 10 of 72 o from 480V Shutdown Board 2A B from 480V Common Board 1 C from 480V Common Board 1 (a) Control air compressors which are powered Obj. V. from the 480 VAC shutdown boards are Obj. V. tripped automatically due to: under voltage on the shutdown boar II. load shed logic during an accident signal concurrent with a loss of offsite powe NOTE: The compressors must be restarted manually after power is restored to the boar (b) Units powered from common boards also trip due to under voltag (4) Lubrication provided from attached oil system via gear-type oil pump (a) Compressor trips on Obj. V. lube oil pressure < 10 psig Obj. V.C.2.
( or lube oil temperature >180 of Ob Ob V.E.12 VD.10 (b) Compressor cylinder is a non lubricated type (5) Cooling water is from the Raw Cooling Water system with backup from EECW (a) Compressor oil cooler, compressor inter-cooler, after cooler and cylinder water jackets (b) Compressor inter-cooler and after cooler moisture traps drain moisture to the Unit 1 station sum NOTE: Cooling water flows to the compressors are regulated Obj. V. such that the RCW outlet temperature is maintained Obj. V. between 70° F and 100° F. Outlet temperatures Obj. V.E.12 should be adjusted low in the band (high flow rates) during warm seasons (river temps . .:::. 70°F). Outlet temperatures should be adjusted high in the band during the cooler seasons (river temps.::: 70°F) to reduce condensation in the cylinder (c) Compressor auto trips if discharge temperature of air> 310° b. Unloaders Obj. VD.10
OPL 171.054 Revision 12 Page 14 of 72 (b) Should both the primary and the backup Cutout switch setpoints controllers fail, all four compressors will come are set at 112 psig to on line at full load until these pressure prevent spurious switches cause the compressors to unload at operation when G air 112psi compressor running (c) When air pressure drops below the high nrp~~1 1- - - Jrp
- - -_. -
r.1 Jt()ff
- -- - - ,- ~ptn()int - - -1- - _. - - (11 n R \ - - - - -
n~i(1) 1- - - v; I thp
-_.-
compressors will again come on line at full load until the high pressure cutoff switches cause the compressors to unload.
d. Relief valves on the compressors discharge set at 120 psig protects the compressor and piping.
e. G Air Compressor - centrifugal type, two stage (1) Located 565' EL Turbine Bldg., Unit 1 en Control Air Compressor G is the primary control air compressor and provides most of the control air needed for normal plant operatio (2) Rated at 1440 SCFM @ 105 psi (3) Power Supply (a) 4 kV Shutdown Board B supplies power to the compressor moto (b) 480 V RMOV Bd. 2A Supplies the following:
* Pre lube pump * Oil reservoir heater * Cooling water pumps * Panel(s) control power * Auto Restart circuit (c) Except for short power interruptions on the 480v RMOV Bd, Loss of either of these two power supplies will result in a shutdown of the G air compresso (4) A complete description of the G Air compressor Cover 01 illustrations controls and indications can be found in 0-01-3 (The G and the F air compressor indications and Microcontrollers are similar).
(a) UNLOAD MODULATE AUTO DUAL TP-8 handswitch is used to select the mode of operation for the compressor
OPL 171.054 Revision 12 Page 30 of 72 3. Component Description Obj. V. a. Compressors E and F (EL 565, U-3 Turbine Building) Obj. V are designated for service ai b. The F air compressor is rated for approximately 630 SCFM @ 105 psig, centrifugal type, 2 stages c. The power supply for both compressors is 480VAC Common Board d. FIG air compressor comparison (1) Controls are similar to that of the G air TP-16 compressor. There is no 4KV breaker control on ObjV. the F air compressor control pane Obj. V. (2) Control system modulates discharge air pressure Set to control at appro in the same manner as is done on the G air 95 psig - Relief Valve is compresso set to lift at ~ 115 psi (3) Air system is similar to the G air compressor. A difference is that the 2 stages of compression are TP-17 driven by one shaft for the F air compressor. On the G air compressor, there is a separate drives; one for each of 3 compression stage (4) Oil system similar to that on the G air compressor TP-18 with exception of location of components and capacity. E compressor has an electric oil pump that runs whenever control power is o (5) Cooling system is similar to that on the G air TP-19 compressor with exception of flow rate, location, and capacity of component (6) Loss of power will result in F air compressor trip, loss of the pre lube pump, and the cooling water pump (7) Restart of the compressor can be accomplished once the compressor has come to a full stop and any trip conditions cleared and rese e. AlarmslTrips (1) The Alert and Shutdown setpoints for the Fair See for latest setpoints compressor are listed in 0-01-3 . RO 300000K3.0 1 OOllelAlT2G lISGTIB lOB/300000K3. 0 1/3 .2/3 .4IRO/SRO/BANK A LOCA has occurred on Unit 1 and the Drywell is being vented to the Standby Gas Treatment System (SBGT), when a loss of the Control Air System occur Which ONE of the following describes the impact of the loss of air, on the operation of DRYWELL VENT INBD ISOL VALVE (1-FCV-64-29), and PATH B VENT FLOW CONT VALVE (1-FCV-S4-19)? Both vent valves 1-FCV-64-29 & 1-FCV-S4-19 will fail CLOSED - - - - - - - - A but, will automatically re-open as soon as INST GAS SELECTOR VALVE (1-PCV-S4-33) automatically swaps to CAD Nitrogen Storage Tank 'A' and must be manually re-opened as soon as INST GAS SELECTOR VALVE (1-PCV-S4-33) automatically swaps to CAD Nitrogen Storage Tank 'A' C.V' but, will automatically re-open as soon as CAD SYSTEM A N2 SHUTOFF VALVE (0-FCV-S4-5) is manually opened from Main Control Room Panel 1-9-5 and must be manually re-opened as soon as CAD SYSTEM A N2 SHUTOFF VALVE (0-FCV-S4-5) is manually opened from Main Control Room Panel 1-9-5 KIA Statement: 300000 Instrument Air K3.01 - Knowledge of the effect that a loss or malfunction of the (INSTRUMENT AIR SYSTEM) will have on the following: Containment air syste KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the effect on the containment air system due to a loss of Control Ai References: 1-EOI Appendicies SG and 12, 1-AOI-32-2 P&L 3. Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam Page: 10f2 2120/2008
REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Whether the vent valves automatically swap to be supplied by CAD/Nitrogen tank 'A' or must be manually aligne . Whether the CAD supply to OW Control Air automatically swaps or must be manually aligne . Failure positions of the Air-Operated Vent Valves 1-FCV-84-19 and 1-FCV-64-2 C - correct: On a loss of Control Air, INST GAS SELECTOR VALVE (1-PCV-84-33) automatically repositions to align CAD tank 'A' (nitrogen) supply to 1-FCV-84-19 and 1-FCV-64-29; to allow venting containment (if necessary, during this event). Even though the 84-33 valve automatically re-aligns, there is a manual operation that must take place to actually align Nitrogen from CAD Tank 'A' to the supply line itself. This manual operation is the opening of CAD SYSTEM 'A' N2 SHUTOFF VALVE (O-FCV-84-5) from Control Room Panel 1-9-54; which is in an odd location (around the back side of the panel and often overlooked). Once this manual action is complete, then and only then will the Air-Operated Vent Valves 1-FCV-84-19 and 1-FCV-64-29 re-open (automatically) to re-establish the vent path from containment. This also involves a recent plant modification, wherein air compressors were previously tied into this air supply line and maintained it pressurized. The air compressors became obsolete, from a parts standpoint, and were thus removed from the system. Now, CAD is the ONLY supply to pressurize the line when Control Air is los A - incorrect: This is plausible because the vent valves in question will automatically re-open AND the INST GAS SELECTOR VALVE (1-PCV-84-33) does, in fact, automatically swaps to CAD Nitrogen Storage Tank 'A.' This centers upon the candidate identifying the impact of the recent plant modification in combination with the potentially obscure operation of having to manually open CAD SYSTEM 'A' N2 SHUTOFF VALVE (O-FCV-84-5) from Control Room Panel 1-9-5 B - incorrect: This is plausible because the vent valves in question will fail closed (and are required to be in the OPEN position to continue venting containment) AND the INST GAS SELECTOR VALVE (1-PCV-84-33) does, in fact, automatically swaps to CAD Nitrogen Storage Tank 'A.' This centers upon the candidate identifying the impact of the recent plant modification, the operation of having to manually open CAD SYSTEM 'A' N2 SHUTOFF VALVE (O-FCV-84-5), and the vent valve re-positioning scheme D - incorrect: This is plausible because the vent valves in question will fail closed (and are required to be in the OPEN position to continue venting containment) AND the potentially obscure operation of having to manually open CAD SYSTEM 'A' N2 SHUTOFF VALVE (O-FCV-84-5) from Control Room Panel 1-9-5 Notes - Changed question substantially for plausibility purposes. Agree LOK is memory.
Page: 2of2 212012008
BFN Loss Of Control Air 1-AOI-32-2 Unit 1 Rev. 0001 Pa~e 5 of 27 SYMPTOMS (continued)
* REACTOR CHANNEL A(B) AUTO SCRAM annunciator, (1-XA-55-5B, Window 1(2)) in alar * MAIN STEAM RELIEF VLV AIR ACCUM PRESS LOW annunciator, (1-XA-55-3D, Window 18) in alarm. AUTOMATIC ACTIONS U-1 TO U-2 CONT AIR CROSSTIE, 1-PCV-032-3901, will CLOSE to separate Units 1 & 2 when control Air Header Control Air Header pressure reaches 65 psig lowering at the valv UNIT 2 TO UNIT 3 CONTROL AIR CROSSTIE, 2-PCV-032-3901, will CLOSE to separate Units 2 and 3 when Control Air Header pressure reaches 65 psig lowering at the valv CAD SUPPLY PRESS REGULATOR, 1-PCV-084-0706, will select nitrogen from CAD Tank A at :::; 75 psig Control Air pressure to supply the following: SUPPR CHBR VAC RELIEF VALVE, 1-FSV-064-0020 SUPPR CHBR VAC RELIEF VALVE, 1-FSV-064-0021 INST GAS SELECTOR VALVE, 1-PCV-084-0033, will select nitrogen from CAD Tank A to supply the following: DRYWELL OR SUPPRESS CHMBR EXHAUST TO SGTS, 1-FSV-084-0019 DRYWELL VENT INBD ISOL VALVE, 1-FSV-064-0029 SUPPR CHMBR VENT INBD ISOL VALVE, 1-FSV-064-0032 INST GAS SELECTOR VALVE, 1-PCV-084-0034, will select nitrogen from CAD Tank B to supply the following: DRYWELL OR SUPPRESS CHMBR EXHAUST TO SGTS, 1-FSV-084-0020 DRYWELL INBD ISOLATION VLV, 1-FSV-064-0031 SUPPR CHBR INBD ISOLATION VLV, 1-FSV-064~003 BFN Loss Of Control Air 1-AOI-32-2 Unit 1 Rev. 0001 Page 7 of 27 Subsequent Actions (continued)
NOTE CNDS BSTR PMPS DISCH BYPASS TO COND 1C, 1-FCV-002-0029A and CNDS BSTR PMPS DISCH BYPASS TO COND 1 B, 1-FCV-002-0029B both fail CLOSED on a loss of control ai [3] IF there is NOT a flow path for Condensate system, THEN STOP the Condensate Pumps and Condensate Booster Pumps. REFER TO 1-01- o
[4] IF any Outboard MSIV closes, THEN PLACE the associated handswitch on Panel 1-9-3 in the CLOSE positio o NOTE RSW STRG TNK ISOLATION, 0-FCV-25-32, fails CLOSED on loss of control ai [5] START a High Pressure Fire Pump. REFER TO 0-01-2 o [6] OPEN CAD SYSTEM A N2 SHUTOFF VALVE, 0-FCV-84-5, at Panel 1-9-5 o [7] OPEN CAD SYSTEM B N2 SHUTOFF VALVE, 0-FCV-84-16, at Panel 1-9-5 o [8] CHECK RCW pump motor amps and PERFORM Steps 4.2[8.1] through 4.2[8.5]to reduce RCW flow:
To 64_20/21 From Air Operators Control Air r------------~~ To Unit 1 (Capped)
$t./--J1 -----l~.I4'4L~
To 84-19 & I
- -@)---§ 64-29n2 ~ From Control Air Air Ope'ratdrs-I
-0I W To Drywell o Standby Gas Control o Treatment AirSystem~ z From ~ CAD N, Tank B z To 84~20 & s:: 64-31/34 m @}-@-,- - -@)---§ Air Operators z-I >> -I s:: 64-30 64-18
~11 ~;t------.-
o 84-20 CJ) -0 ~_ _~I___ 11~ I m
- 0 I
i-----~ I~f-m 64-140 o r C :DRYWELL I "." ,,~, -I o I DELTAP i I COMP I 64-141 Z I I CJ) I I 64-19 L _____ ~ ... 11~ ~ 64-34 CJ) I~t-I---- -I 64-139 ! m s::
-0>>;;00 Q) -0 CD -0 CO-o<r CD CD 00" .....
:::J -"-....J Wc.. 0 .....
CD -":::J OX .....
-'ONW o
~ N CO
1-EOI APPENDIX-12 BFN PRIMARY CONTAINMENT VENTING Rev. 0 UNIT 1 Page 4 of 8 VERIFY 1-FIC-84-20, PATH A VENT FLOW CONT, is indicating approximately 100 scf CONTINUE in this procedure at step 12.
1 VENT the Drywell using 1-FIC-84-19, PATH B VENT FLOW CONT, as follows: VERIFY CLOSED 1-FCV-64-141, DRYWELL DP COMP BYPASS VALVE (Panel 1-9-3). PLACE keylock switch 1-HS-84-36, SUPPR CHBR/DW VENT ISOL BYP SELECT, to DRYWELL position (Panel 1-9-54). VERIFY OPEN 1-FCV-64-29, DRYWELL VENT INBD ISOL VALVE (Panel 1-9-54). PLACE 1-FIC-84-19, PATH B VENT FLOW CONT, in AUTO with setpoint at 100 scfm (Panel 1-9-55). PLACE keylock switch 1-HS-84-19, 1-FCV-84-19 CONTROL, in OPEN (Panel 1-9-55). VERIFY 1-FIC-84-19, PATH B VENT FLOW CONT, is indicating approximately 100 scf CONTINUE in this procedure at step 12.
1 VENT the Drywell using 1-FIC-84-20, PATH A VENT FLOW CO NT, as follows: VERIFY CLOSED 1-FCV-64-1A 1, DRYWELL DP COMP BYPASS VALVE (Panel 1-9-3). PLACE keylock switch 1-HS-84-35, SUPPR CHBR lOW VENT ISOL BYP SELECT, to DRYWELL position (Panel 1-9-54). VERIFY OPEN 1-FCV-64-31, DRYWELL INBD ISOL VALVE (Panel 1-9-54). VERIFY 1-FIC-84-20, PATH A VENT FLOW CONT, in AUTO with setpoint at 100 scfm (Panel 1-9-55). PLACE keylock switch 1-HS-84-20, 1-FCV-84-20 ISOLATION BYPASS, in BYPASS (Panel 1-9-55). VERIFY 1-FIC-84-20, PATH A VENT FLOW CONT, is indicating approximately 100 scf cm Z"Tl TO REACTOR BLDG VENT SYSTEM OVERVIEW =i Z EXHAUST FANS .....
~I i Y\Oi I- - ~,------ -- ----*-----A---- -1 64 -40 1 I f -----.)-- TO SGTS i
I I" 64-41 i Ii
-~-- I iA~'
L-.~
'I ID-0
:::0 S
~1j I--~
64-30 T -[-- ---,- - i 64-29
-~
18"
--T ---- -- ___
>>
:::0 iREi r;;a I ~OI J n <
[9 I I I (')
~ 2*I~J¥g~r~'2:'~
(Xl (Xl
"'iRE' 0 I DRYWELL Z
--I
>>
. I I ~3 " Z L-_-
i
~CJd4:1~0 ~ .
S m Z I L,p ~I-~ --I I < L_'~:-i1f: __I' COM ::: - DRYWELL PURGE FANS AND FILTERS m Z
~
I :::!
---.----- .
II V'
~--: 1r--O--i-~ J ~~ ~ z i I - - G)
I ~! I I_ _ ~ ___,___
~~ ,)~cj
__~, ___ ... L _________ 64-33 64-32 ..... I REACTOR BLDG m VENTILATION EXHAUST >>
--I ->>0
--I
>>-0 -0 (') Il) -0
::J: CQ m S (I) Z m-.....:::o!2 Zo(l)><
--1_:C:~
..... 0)01>.)
1-EOI APPENDIX-8G BFN CROSSTIE CAD TO Rev. 0 UNIT 1 DRYWELL CONTROL AIR Page 1 of 2 LOCATION: Unit 1 Control Room ATTACHMENTS: None OPEN the following valves:
* 0-FCV-84-5, CAD A TANK N2 OUTLET VALVE (Unit 1, Panel 1-9-54) * 0-FCV-84-16, CAD B TANK N2 OUTLET VALVE (Unit 1, Panel 1-9-55). VERIFY O-PI-84-6, VAPOR A OUTLET PRESS, and 0-PI-84-17, VAPOR B OUTLET PRESS, indicate approximately 100 psig Panel 1-9-54 and Panel 1-9-55). PLACE keylock switch 1-HS-84-48, CAD A CROSS TIE TO OW CONTROL AIR, in OPEN (Panel 1-9-54). CHECK OPEN 1-FSV-84-48, CAD A CROSS TIE TO OW CONTROL AIR, (Panel 1-9-54). PLACE keylock switch 1-HS-84-49, CAD B CROSS TIE TO OW CONTROL AIR, in OPEN (Panel 1-9-55). CHECK OPEN 1-FSV-84-49, CAD B CROSS TIE TO OW CONTROL AIR (Panel 1-9-55). CHECK MAIN STEAM RELIEF VLV AIR ACCUM PRESS LOW, 1-PA-32-31, alarm cleared (1-XA-55-3D, Window 18). IF ............. ;... MAIN STEAM RELIEF VLV AIR ACCUM PRESS LOW, 1-PA-32-31, annunciator is or remains in alarm (1-XA-55-3D, Window 18),
THEN ........... DETERMINE which Drywell Control Air header is depressurized as follows: DISPATCH personnel to Unit 1, RB, EI 565 ft, to MONITOR the following indications for low pressure:
* 1-PI-084-0051, DW CONT AIR N2 SUPPLY PRESS indicator, for CAD A (RB, EI. 565, by Drywell Access Door), * 1-PI-084-0050, DW CONT AIR N2 SUPPLY PRESS indicator, for CAD B (RB, EI. 565, left side of 480V RB Vent Board 1B).
25. RO 400000A2.02 001lCIAlT2GlIRBCCWI1400000A2.02/3.8/4.lIRO/SROINEW 0112012008 Unit 2 is operating at 100% power, and the following annunciators are received:
* DRYWELL EQPT DR SUMP TEMP HIGH (2-XA-55-4C, Window 16) * RBCCW SURGE TANK LEVEL LOW (2-XA-55-4C, Window 13) * RBCCW PUMP DISCH HDR PRESS LOW (2-XA-55-4C, Window 12) * RBCCW 2-FCV-70-48 CLOSED (2-XA-55-4C, Window 19)
Which ONE of the following describes the impact on Reactor Building Closed Cooling Water (RBCCW) System based on these indications, and the correct actions required? RBCCW flow to the RWCU Non-Regenerative Heat Exchanger has been los Enter 2-AOI-70-1, "Loss of RBCCW" and immediately trip the RWCU pump I sal.~ ~ (tlAJ c..u.... s'(1-S:t::.- . RBCCW flow to the Drywell has been los Enter 2-AOI-70-1, "Loss of RBCCW", Scram the Reactor, and immediately trip the Recirc Pump Makeup to the RBCCW Surge Tank in accordance with the Alarm Response Procedure; then, place the Spare RBCCW Pump in service and re-open RBCCW SECTIONALIZING VALVE (2-FCV-70-48).
D." RBCCW flow to the RWCU Non-Regenerative Heat Exchanger has been los Enter 2-AOI-70-1, "Loss of RBCCW" and remove the RWCU pumps from service IF RBCCW suction temperature cannot be maintained below 105°F.
Page: 10f3 2/20/2008
KIA Statement: 400000 Component Cooling Water A2.02 - Ability to (a) predict the impacts of the following on the CCWS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: High/low surge tank level KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the effect of a leak into the RBCCW system and determine which procedure addresses this conditio References: 2-AOI-70-1, 4.2[7], 1-ARP-9-4C rev. 0015 window 12,13,16 & 19 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 outcom NRC Exam MODIFIED FROM OPL 171.047 #9 Page: 20f3 2/20/2008
REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. The impact on RBCCW based on given condition . What actions would be required to mitigate the proble D - correct: With the given alarms it can be deduced that there has been a loss of the RBCCW non-essential loop. The appropriate action would be to enter 2-AOI-70-1 and shutdown the RWCU pumps (when the RBCCW suction temperature exceeds 105°F).
A - incorrect: RBCCW has been lost to the Non-regenerative Heat Exchanger but it is not appropriate to immediatley trip the RWCU pumps. This is plausible due to the immediate actions in 2-AOI-70-1 which have the operator trip the RWCU pumps in the event an RBCCW trips and can't be restore B - incorrect: It cannot be determined from the given conditions that Orywell cooling has been lost. 2-FCV-70-47 isolates the Orywell; NOT 2-FCV-70-48. This is plausible because the given conditions indicate the leak is possibly located inside the orywe II , however it is not appropriate to Scram the Reactor and trip the Recirc pumps at this tim C - incorrect: This is plausible because all of the actions are inaccordance with approved plant procedures but placing the Spare pump inservice is not applicable for this even Note - rewrote the stem and 3 of the distractors.
Page: 30f3 2120/2008
BFN Loss of Reactor Building Closed 2-AOI-70-1 Unit 2 Cooling Water Rev. 0027 ( Page 8 of 14 Subsequent Actions (continued) NOTE Opening RBCCW TVC Bypass valves may cause an EECW pump start due to low pressur [7] IF RBCCW PUMP SUCTION HDR TEMP, 2-TIS-70-3 (Panel 2-9-4), cannot be maintained below 105°F, THEN PERFORM the following (otherwise N/A):
* SLOWLY OPEN bypass valves around RBCCW TCV o * PLACE Spare RBCCW Heat Exchanger in servic REFER TO 2-01-7 o * VERIFY RWCU System removed from servic o (
NOTE It may be necessary to place the RHR System in the Supplemental Fuel Pool Cooling Mode to maintain Fuel Pool temperature below 150°F (Tech. Specs. 3.1 0.C.2, TRM 3.9.2).
- SHUT DOWN Fuel Pool Cooling System. REFER TO 2-01-7 o
BFN Panel 9-4 2-ARP-9-4C Unit 2 2-XA-55-4C Rev. 0027 PC:lg~ 19 of 44 SensorlTrip Point: RBCCWPUMP 2-PS-70-15 50 psig (57 psig on local pump discharge DISCH. HDR pressure gauge) Press LOW 2-PA-70-15 (Page 1 of 1) Sensor Panel 25-8 Location: Elevation 593 Column R-13 R-UNE Probable A. Failed pum Cause: B. Broken couplin C. Sensor malfunctio Automatic Closes 2-FCV-70-48, non-essential loop, closed cooling water sectionalizing MOV, Action: ( Operator VERIFY 2-FCV-70-48 CLOSING/CLOSE D Action: VERIFY RBCCW pumps A and B in servic D VERIFY RBCCW surge tank low level alarm is rese D DISPATCH personnel to check the following: D
* RBCCW surge tank level locall D * RBCCW pumps for proper operatio D E. REFER TO 2-AOI-70-1 for RBCCW System failure and 2-01-70, for starting spare pum D References: 45N620-4 45N779-6 2-47E61 0-70-1 FSAR Sections 10.6.4 and 13. BFN Panel 9-4 2-ARP-9-4C Unit 2 2-XA-55-4C Rev. 0027 Page 20 of 44 SensorlTrip Point:
RBCCW 2-LS-70-2B 4 inches below center line of tank SURGE TANK LEVEL LOW 2-LA-70-2B (Page 1 of 1) Sensor On the RBCCW surge tank in the MG Set Room, EI 639' Location: Probable A. Normal leakage.
Cause: B. Drain valves ope C. Abnormal leakage.
Automatic None Action: Operator A. ADD water to the RBCCW Surge Tank for approximately one minute Action: or until low level alarm resets using the following: .0
* RBCCW SYS SURGE TANK FILL VLV, 2-FCV-70-1 (Panel 2-9-4) OR 0 * FCV-70-1 BYPASS VLV, 2-HCV-2-1369 (locally). 0 B. IF alarm does NOT resets, THEN CHECK tank locall C. IF unable to maintain RBCCW Surge Tank level, THEN REFER TO 2-AOI-70- D. IF necessary to add water more than once per shift, THEN CHECK Drywell floor drain system for excessive operation AND INSPECT system outside Drywell for leakag References: 45N614-4 2-47E61 0-70-1 45N620-4 FSAR Sections 10.6.4 and 13. BFN Panel 9-4 2-ARP-9-4C Unit 2 2-XA-55-4C Rev. 0027 Page 23 of 44 Sensor/Trip Point:
DRYWELL 2-TS-077-0014 EQPT DR SUMP TEMP HIGH 2-TA-77-14 (Page 1 of 1) Sensor Elevation 550 at Drywell equipment drain sump. 2-TM-77-14 on Panel 9-19 in Location: Auxiliary Instrument Room.
Probable A. Recirculation loop A or B drain and miscellaneous reactor drains have excessive Cause: leakag B. Loss of RBCC C. Reactor refueling bellows excessive leakag D. CRD equipment service drain excessive leakage.
Automatic A. Starts sump pump.
Action: B. Opens recirculation valve 2-FCV-77-14A and closes discharge valve 2-FCV-77-14B (on Panel 2-9-4).
Operator Action: NOTE DRYWELL EQPT DRAIN SUMP TEMP, 2-TIS-77-14 pump shutoffsetpoint (Blue Pen) may be adjusted in the range of 115 to 125 degrees F. Adjustment of this setpoint with respect to seasonal changes in river temperature may minimize Drywell Equipment Drain Sump Recirculation Mode initiation frequency and/or minimize pump run tim A. CHECK OPEN 2-FCV-77-14A o B. CHECK CLOSED 2-FCV-77-14 o C. CHECK Drywell equipment drain sump pump runnin o D. VERIFY temperature lowers and alarm reset o E. IF RBCCW has been lost, THEN REFER TO 2-AOI-70- o F. IF alarm does NOT resets after 15 minutes, THEN NOTIFY Shift Manage o References: 45N620-4 2-47E61 0-77-1 2-47E852-2 FSAR Sections 10.16.6 and 13. BFN Panel 9-4 2-ARP-9-4C Unit 2 2-XA-55-4C Rev. 0027 ! L ... _ .................._ _ _ ......._ ......... _ _ .,............._ ........ _ .. __ .*.. _ ... _ ............ ~ ..*. _ ............ _ ... _ .*. _ ..._ _ ... _ . _ ... _ ... ~_~g~. ~~_<?f 4~L .._. . ___. . __._ .._J Sensor/Trip Point: RBCCW Limit Switch on 2-FCV-70-48 2-FCV-70-48 FULLY CLOSED 2-FCV-70-48 CLOSED 2-ZA-70-48 (Page 1 of 1) Sensor Elevation 593' Location: RBCCW System routing valve to the nonessential loop.
Probable A. Manual operation of valve handswitch.
Cause: B. Low pump discharge header pressure:s; 50 psi C. Re-energizing after a loss of normal AC power in conjunction with accident signal on affected or adjacent uni D. Failed pressure switch, 2-PS-070-0015, RBCCW PUMP DISCH HDR.
Automatic None Action: Operator A. IF cause is accident signal AND loss of offsite power, THEN Action: REFER TO 0-AOI-57-1A and 2-EOI-1 FLOWCHART and 2-EOI-2 FLOWCHAR o B. OPEN 2-FCV-70-48, RBCCW Sectionalizing Valve, when conditions p~m~ 0 C. IF unable to reopen 2-FCV-70-48, THEN if desired, REMOVE RWCU from service. REFER TO 2-01-6 . NOTIFY CHEMISTRY if RWCU is removed from service (Reference TRM 3.4.1). 0 D. REFER TO 2-AOI-70- References: 2-47E61 0-70-1 45N779-6 45N620-4 FSAR Sections 10.6.4 and 13. . RO 400000G2.4.31 001lCIAlT2GIIRBCCWI14000002.4.30//RO/SRO/NEW 10116107 Unit 3 is at rated power with the following indications:
* RECIRC PUMP MTR 'B' TEMP HIGH (3-XA-9-4B, Window 13).
- RBCCW EFFLUENT RADIATION HIGH (3-XA-9-3A, Window 17).
- RBCCW SURGE TANK LEVEL HIGH (3-XA-9-4C, Window 6).
- RX BLDG AREA RADIATION HIGH (3-XA-9-3A, Window 22).
- Recirc Pump Motor '3B' Winding and Bearing Temperature Recorder (3-TR-68-84) is reading 1700 F and risin * RBCCW Pump Suction Header Temperature Indicator (3-TIS-70-3) is reading 104°F and risin * RWCU NON-REGENERATIVE HX DISCH TEMP HIGH (3-XA-9-4C, Window 17).
- Area Radiation Monitor RE-90-13A and RE-90-14A are in alarm reading 55 mr/hr and risin Which ONE of the following describes the action(s) that should be taken in accordance with plant procedures?
REFERENCE PROVIDED Enter 3-EOI-3, "Secondary Containment Control" and _ _ _ _ _ _ _ _ _ __ A.'" Trip and isolate '38' Recirc Pum Enter 3-AOI-68-1A "Recirc Pump Trip/Core Flow Decrease OPRMs Operable." Trip and isolate '3B' Recirc Pum Immediately commence a shutdown in accordance with 3-GOI-1 00-12A, "Unit Shutdown." Trip RWCU Pumps and isolate the RWCU syste Commence a normal shutdown in accordance with 3-GOI-1 00-12A, "Unit Shutdown." Trip RWCU pumps and isolate RWCU syste Close RBCCW Sectionalizing Valve 3-FCV-70-48 to isolate non-essential loads and maximize cooling to '3B' Recirc Pump.
Page: 1 of 3 2/20/2008
KIA Statement: 400000 Component Cooling Water 2.4.31 - Emergency Procedures I Plan Knowledge of annunciators alarms and indications, and use of the response instruction KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the corrective actions required due to an emergency involving RBCCW based on annunciators and indication References: 3-EOI-3 flowchart, 3-AOI-68-1A rev 4, 4.2[8.4], 3-ARP-9-3A rev 36 window 17 and 3-ARP-9-4B rev 36 window 13 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 outcom NRC Exam ( Page: 2of3 2120/2008
REFERENCE PROVIDED: 3-EOI-3 flowchart Plausibility Analysis: In order to answer this question correctly, the candidate must determine the following: 1. EOI Entry is required solely based on Area Radiation Monitor Alarm . Location of the leak is from the 3B Recirc Pum . RWCU temperature indications are due to insufficient cooling by RBCCW, not a RWCU lea . Appropriate actions per 3-EOI-3 are to isolate the leak and monitor radiation level . Justification for Unit Shutdown and Cooldown are due to the Recirc Loop being isolated at rated temperature and pressure (pipe hanger and support issue), and NOT Directed by 3-EOI- A - correct: EOI-3 entry is required due RE-90-13A & 14A in alarm. Once the 3B Recirc pump is isolated the appropriate action is to enter 3-AOI-68-1A which will eventually require the unit to be shutdown lAW GOI-100-12A due to not being able to maintain the required delta T between the isolated recirc loop and the steam dome temperatur B - incorrect: An immediate shutdown is not required at this time. See 'A'. This is plausible because the location of th.e leak and requirement for isolation are correc C - incorrect: This is plausible if the candidate incorrectly determines that RWCU is causing the temperature issues with 3B Recirc Pump and not vice versa. In addition to the justification above, commencing a shutdown in accordance with 3-EOI-3 is not appropriate until ARMs indicate greater than 1000 mr/h D - incorrect: This is plausible if the candidate incorrectly determines that RWCU is causing the temperature issues with 3B Recirc Pump and not vice versa. If RWCU was the leak location, the RBCCW temperature would NOT be high enough to provide the given indications. The leak would have to have occurred in the NRHX which is below the indicated RBCCW temperatur Note - changed the RBCCW suction temp from 1400 F to 104°F. A manual Scram is required at 1100 F per AOI-70-1. Also changed distractor 'B' to enter AOI-68-1A which would be the correct actions under the condition Changed the Stem, took Enter 3-EOI-3, "Secondary Containment Control" out of the distractors and put it in the Stem.
Page: 3 of3 2/20/2008
OPL 171.047 Revision 12 Appendix C Page 35 of 41 DEMIN WATER-~-I MAKEUP DRW RCW OUTLET RBCCW RETURN'-~~~--J RCW HEADER TCV'S U2-1~H U3-1~i-I 0-70-607 CHEMICAL FEED RCW U2 TCV'S U3 RBCCW SUPPLY HEADER 67 69 U2 U3 68 70 TP-1: RBCCW SYSTEM FLOW DIAGRAM
OPL 171.034 Revision 11 Appendix C Page 30 of 30 EOI- 3 TABLE 4 SECONDARY CONTAINMENT AREA RADIATION APPLICABLE MAX NORMAL MAX SAFE POTENTIAL AREA RADIATION VALUE VALUE ISOLATION INDICATORS MR/HR MR/HR SOURCES RHR SYS I PUMPS 90-25A ALARMED 1000 FCV-74-47,48 RHR SYS II PUMPS ALARMED 1000 FCV-74-47,48 90-28A HPCI ROOM FCV-73-2, 3, 81 90-24A ALARMED 1000 FCV_73-44 CS SYS I PUMPS 1000 90-26A ALARMED FCV-71-2, 3, 39 RCIC ROOM CS SYS II PUMPS 90-27 A ALARMED 1000 NONE FCV-73-2, 3, 81 TORUS ALARMED 1000 90-29A FCV-74-47,48 GENERAL AREA FCV-71-2,3 RB EL 565 W 90-20A ALARM ED 1000 FCV-69-1, 2,12 SDV VENTS & DRAINS RB EL 565 E 90-21A ALARMED 1000 SDV VENTS & DRAINS RBEL565NE ALARMED 1000 NONE 90-23A TIP ROOM 90-22A ALARMED 100,000 TIP BALL VALVE RB EL 593 ALARMED 1000 FCV-74-47,48 90-13A,14A RB EL621 ALARMED 1000 FCV-43-13, 14 90-9A RECIRC MG SETS 90-4A ALARMED 1000 NONE REFUEL FLOOR 90-1A, 2A, 3A ALARMED 1000 NONE TP -7 EOI-3 TABLE 4
BFN Recirc Pump Trip/Core Flow Decrease 3-AOI-68-1 A Unit 3 OPRMs Operable Rev. 0004 Page 8 of 16 Subsequent Actions (continued) CAUTION The temperature of the coolant between the dome and the idle Recirc loop should be maintained within 75°F of each other. If this limit cannot be maintained, a plant cool down should be initiated. Failure to maintain this limit and not cool down could result in hangers and/or shock suppressers exceeding their maximum travel range. [GE SIL 251,430 and 517]
[8] IF Recirc Pump was tripped due to dual seal failure, THEN (Otherwise N/A) [8.1] VERIFY TRIPPED, RECIRC DRIVE 3A(3B) NORMAL FEEDER, 3-HS-57-17(14). D [8.2] VERIFY TRIPPED, RECIRC DRIVE 3A(3B)
ALTERNATE FEEDER, 3-HS-57-15(12). D
[8.3] CLOSE tripped recirc pump suction valve using, RECIRC PUMP 3A(3B) SUCTION VALVE, 3-HS-68-1 (77). D [8.4] IF it is evident that 75°F between the dome AND the idle Recirc loop cannot be maintained, THEN COMMENCE plant shut down and cool down. Refer to 3-GOI-100-12 D [9] NOTIFY Reactor Engineer to PERFORM the following: * Tech Specs 3. D * 3-SR-3.4.1 (SLO), Reactor Recirculation System Single Loop Operation D * 0-TI-248, Core Flow Determination in Single Loop Operation D [10] [NER/C] WHEN the Recirc Pump discharge valve has been closed for at least five minutes (to prevent reverse rotation of the pump) [GE SIL-517], THEN (N/A if Recirc Pump was isolated in Step 4.2[8])
OPEN Recirc Pump discharge valve as necessary to maintain Recirc Loop in thermal equilibriu D
BFN Panel 9-3 3-ARP-9-3A Unit 3 3-XA-55-3A Rev. 0036 P-age 25 of 51 Sensor/Trip Point: RBCCW EFFLUENT RADIATION HI-HI HIGH
.t!!
RE-90-131D (NOTE 2) (NOTE 2) 3-RA-90-131 A Hi alarm from recorder Hi-Hi alarm from drawer (Page 1 of 2)
(2) Chemlab should be contacted for current setpoints per O-TI-45.
Sensor RE-90-131A RBCCW HX Rx Bldg, EI 593, R-20 S-LiNE Location: Probable HX tube leak into RBCCW system.
Cause: Automatic None Action: Operator A. DETERMINE cause of alarm by observing following: Action: 1. RBCCW and RCW EFFLUENT RADIATION recorder, 3-RR-90-131/132 Red pen on Panel 3-9- D 2. RBCCW EFFLUENT OFFLINE RAD MON, 3-RM-90-131D on Panel 3-9-1 D B. NOTIFY Chemistry to sample RBCCW for total gamma activity to verify conditio D C. START an immediate investigation to determine if source ofleak is RWCU Non-regenerative, Fuel Pool Cooling, Reactor Water Sample or RWCU Recirc Pump 3A or 3B Seal Water heat exchanger(s). D D. [NER/C] CHECK Following for indication of Reactor Recirculation Pump Seal Heat Exchanger leak: 1. LOWERING in reactor Recirculation pump 3A(3B) No.1 or 2 SEAL, 3-PI-68-64A or 3-PI-68-63A (3-PI-68-76A or 3-PI-68-75A) on Panel 3-9- D 2. Temperature rise on CLG WTR FROM SEAL CLG TE-68-54, on RECIRC PMP MTR 3A WINDING AND BRG TEMP temperature recorder, 3-TR-68-58, on Panel 3-9-2 D 3. Temperature rise on CLG WTR FROM SEAL CLG TE-68-67, on RECIRC PMP MTR 3B WINDING AND BRG TEMP temperature recorder, 3-TR-68-84, on Panel 3-9-2 D Continued on Next Page
BFN Panel 9-3 3-ARP-9-3A Unit 3 3-XA-55-3A Rev. 0036 ( Page 26 of 51 RBCCW EFFLUENT RADIATION HIGH 3-RA-90-131A, Window 17 (Page 2 of 2) Operator Action: (Continued) E. IF it is determined the source of leakage is from Reactor Recirc Pump A(B), THEN 1. ISOLATE Reactor Recirculation Loop A(B) per 3-01-68, as applicabl NOTE Cooldown is required to prevent hangers or shock suppressors from exceeding their maximum travel rang . WHEN primary system pressure is less than 125 psig, THEN ISOLATE RBCCW System to preclude damage to RBCCW piping. [lEN 89-054, GE SIL-459] o References: 3-45E620-3 3-4 7E61 0-90-3 GE 3-729E814-3
BFN Panel 9-4 3-ARP-9-4B Unit 3 3-XA-55-4B Rev. 0036 Page 17 of 45 SensorlTrip Point: Alarm is from 3-TR-68-84, Panel 3-9-2 RECIRC 3-TE-68-73A RECIRC PMP MTR 3B-THR BRG UPPER FACE (190°F) PUMP MTR B 3-TE-68-73C RECIRC PMP MTR 3B-THR BRG LOWER FACE (190°F) TEMP HIGH 3-TE-68-73E RECIRC PMP MTR 3B-UPPER GUIDE BRG (190°F) 3-TA-68-84 3-TE-68-73N RECIRC PMP MTR 3B-LOWER GUIDE BRG (19DOF) 3-TE-68-73G RECIRC PMP MTR 3B-MOTOR WINDING A (216°F) 3-TE-68-73J RECIRC PMP MTR 3B-MOTOR WINDING B (216°F)
(Page 1 of 1) 3-TE-68-73L RECIRC PMP MTR 3B-MOTOR WINDING C (216°F)
3-TE-68-73T RECIRC PMP MTR 3B-SEAL NO.2 CAVITY(180°F) 3-TE-68-73U RECIRC PMP MTR 3B-SEAL NO.1 CAVITY(180°F) 3-TE-68-67 RECIRC PMP MTR 3B-CLG WTR FROM SEAL CLG (140°F) 3-TE-68-70 RECIRC PMP MTR 3B-CLG WTR FROM BRG (140°F) Sensor Temperature elements are located on recirculation pump motor, Elevation 563.12, Location: Unit 3 drywell.
Probable A. Possible bearing failure.
Cause: B. Possible motor overloa C.lnsufficient cooling wate D. Possible seal failur E. High drywell temperature.
Automatic None Action: Operator A. CHECK following on Panel 3-9-4: D Action: * RBCCW PUMP SUCTION HDR TEMP temperature indicating switch, 3-TIS-70-3 normal (summer 70-95°F, winter 60-80°F). D
* RBCCW PRI CTMT OUTLET handswitch, 3-HS-70-47A (3-FCV-70-47) OPE D B. CHECK the temperature of the cooling water leaving the seal and bearing coolers < 140°F on RECIRC PMP MTR 3B WINDING AND BRG TEMP temperature recorder, 3-TR-68-84 on Panel 3-9-2 D C. LOWER recirc pump speed until Bearing and/or Winding temperatures are below the alarm setpoin D D. CONTACT Site Engineering to PERFORM a complete assessment and monitoring of all seal conditions particularly seal leakage, temperature, and pressure of all stages for Recirc Pump seal temperatures in excess of 180° D References: 3-45E620-5 3-4 7E61 0-68-1 Tech Spec 3. GE 731 E320RE 3-SIMI-68B FSAR Section 13. (
EXAMINATION REFERENCE
.PROVIDED TO CANDIDATE
CW) . -w
~ g * * ~
o I**
~ ~
m~
~~. ~i~ ~o i~ i C") -ow*
27. RO 201003K3.03 001IMEMlT2G2/85-31B111201003K3.03/3.6/3.7IRO/SROIBANK Given the following plant conditions:
* AOI 85-3, "CRD System Failure," directs a manual scram based on low reactor pressur Which ONE of the following PROCEDURAL reactor pressure limits should be adhered to in this case and what is the basis for the limit? psig reactor pressur This is the lowest pressure required to move the control rods based on the weight of the control rod blad psig reactor pressur This ensures the Technical Specification scram insertion time limit is not violated in the event of an automatic scra C." 900 psig reactor pressur This would be the lowest pressure a scram can be ensured due to the loss of accumulator psig reactor pressur This would be the lowest pressure a scram can be ensured in the event of a loss of CRD pump KIA Statement:
201003 Control Rod and Drive Mechanism K3.03 - Knowledge of the effect that a loss or malfunction of the CONTROL ROD AND DRIVE MECHANISM will have on following: Shutdown margin KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific knowledge of CRD mechanism limitations and the basis for that limitation related to the ability to effect and maintain shutdown margi References: 1/2/3-AOI-85-3, sec 4.1 and OPL 171.006 rev 9 pg 36, 38 & TP-9 Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam Page: lof2 2/20/2008
REFERENCE PROVIDED: None Plausibility Analysis: ( In order to answer this question correctly the candidate must determine the following: 1. The minimum pressure allowed by 1/2/3-AOI 85-3, CRD System Failur . The basis for that minimum pressur C - correct: <900 psig reactor pressure is the AOI limit for inserting a manual reactor scram with a failure of the CRD system. Below 900 psig (actual pressure is 800 psig) a reactor scram on pressure alone may not be within the design basis requirement A - incorrect: This is plausible because the entire statement is accurate, but is NOT the pressure specified by 1/2/3-AOI 85-3, CRD System Failure and is NOT the correct reason for inserting a scra B - incorrect: This is plausible because the entire statement is accurate, but is NOT the pressure specified by 1/2/3-AOI 85-3, CRD System Failur D - incorrect: This is plausible because 940 psig is the other pressure (charging water pressure) refered to in AOI-85- Note - Changed the Stem to say what is the basis for the limit vs. 'WHY?'. Changed distractor 'A' to say 'to move a control rod blade based on the weight of the blade.' Changed distractor 'B' to say 'ensures the Technical Specification scram insertion time limit is not violated in the event of an automatic scram.' Changed distractor '0' to '940 psig' and 'in the event of a loss of CRD pumps' instead of 'this would be the lowest pressure a scram can be ensured due to the loss of accumulators.' Page: 2of2 2/20/2008
BFN CRD System Failure 3-AOI-85-3 Unit 3 Rev. 0009 Page 7 of 11 Immediate Actions (continued)
[3] IF Reactor Pressure is LESS THAN 900 PSIG AND ANY ONE of the following conditions exist: * In service CRO Pump tripped and NEITHER CRO Pump can be started, OR * Charging Water Pressure can NOT be restored and maintained above 940 PSIG, THEN PERFORM the following: (Otherwise N/A) [3.1 ] MANUALLY SCRAM Reactor, PLACE the reactor mode switch in the shutdown position immediatel o [3.2] REFER TO 3-AOI-100-1. [Item 020] o
OPL171.006 Revision 9 Page 36 of 60 ( Fnet = (Rx Pressure x Under-Piston Area) -
(Rx Pressure x Area of Index Tube Note: 4 in 2 + Weight of Blade + Friction)
upward force - 1.2 in 2 Fnet = (1000 psig x 4.0 in 2) - [1000 psig downward force
= 2.8 in 2 x (4.0 in 2 - 1.2 in 2 )] - 255 Ibs - - 500 Ibs Fnet = 4000 - 2800 - 255 - 500 Fnet = 445 Ibs (Upward) Single failure proof - There is no single-mode failure to the hydraulic system which would prevent the drive from scrammin d. ' AccumulatorVersus reactor vessel pressure scrams (1 ) TP-9 represents a plot of 90 percent TP-9 scram times versus reactor pressur (a) Reactor pressure only (b) Accumulator pressure only (c) Combined reactor and accumulator pressure (2) Scram times are measured for only the first 90% of the rod insertion since the buffer holes at the top end of the stroke slow the driv (3) Reactor-pressure-only scram (a) As can be seen from TP-9, the drive cannot be scrammed with reactor pressure ~ 400 psi (b) The net initial upward force available to scram the drive can be calculated as follow OPL 171.006 Revision 9 Page 38 of 60 e. Average scram times (normal drive) TP-9 (1 ) Technical Specifications state that scram times are to be obtained without reliance on the CRD pump (2) Consequently, the charging water must be valved out on the drive to be teste (3) Maximum scram time fora typical drive occurs at 800 psig reactor pressur (4) This is why Technical Specifications specify that scram times are to be taken at 800 psig or greater reactor pressure.
f. Abnormal scram conditions (1 ) Scram outlet valve failure to open (2) Drive will slowly scram on seal leakage as long as accumulator charging water pressure stays greater than reactor pressur (3) If the accumulator is not available, the drive will not scram (this is a double failure).
g. Control Rods failure to Insert After Scram Obj. V.D.11 (1 ) This condition could be due to hydraulic loc (2) Procedure has operator close the See 2-01-85 & 2-Withdraw Riser Isolation valve. Connect EOI App-1 E for drain hose to Withdraw Riser Vent Test detailed Connection on the affected HCU. Slowly operations open Withdraw Riser Vent. When inward motion has stopped, close Withdraw Self Check Riser Ven Peer Check
OPL 171.006 Revision 9 Appendix C Page 59 of60 .0 .0 ro o z o () LU ACCUMULATOR PRESSURE ONLY ~ (VESSEL PORTS PLUGGED) LU
- 2 (CHARGING WATER VALVED OUT)
i=
- 2
<< 5 3.0 CfJ ~ o OJ .0 o 200 400 600 800 1000 1200 TP-9: Scram Time Versus Reactor Pressure
28. RO 201006K4.09 001lMEM/T2G2IRWMI1201006K4.09/3.2/3.2/RO/SROINEW 11116107 RMS Which ONE of the following describes how the Rod Worth Minimizer (RWM) System ( INITIALIZATION is accomplished? INITIALIZATION occurs automatically when the RWM is unbypasse B ..... INITIALIZATION must be performed manually when the RWM is unbypasse INITIALIZATION occurs automatically every 5 seconds while in the transition zon INITIALIZATION must be performed manually when power drops below the Low Power Setpoint (LPSP).
KIA Statement: 201006 RWM K4.09- Knowledge of ROD WORTH MINIMIZER SYSTEM (RWM) (PLANT SPECIFIC) design feature(s) and/or interlocks which provide for the following: System initialization: P-Spec(Not-BWR6) KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific of which plant condition would INITIALIZE the RW References: 1/2/3-01-85, OPL 171.024 rev 13, pg 19 - f.(2) Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam ( Page: lof2 2/20/2008
REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. When RWM INITIALIZATION is require . How RWM INITIALIZATION is accomplishe B - correct: Initialization must be performed whenever the RWM has been taken off line, as occurs whenever the RWM program is aborted or manually bypasse A - incorrect: This is plausible because initialization is required when the RWM is unbypassed; but, this must be done manuall C - incorrect: This is plausible because the RWM automatically initiates a "scan/latch" to determine the correct latched rod group; but, this is NOT the same as INITIALIZATIO D - incorrect: The RWM does NOT require manual INITIALIZATION when the Low Power Setpoint (LPSP) is reached IF it has NOT been manually bypassed or had a program abor Notes - Changed distractors by removing the phrase 'using the INITIALIZATION push button'. Which button/buttons are depressed is irrelevant because it will be performed with the procedure in hand.
Page: 2of2 2120/2008
OPL 171.024 Revision 13 Page 19 of 53 INSTRUCTOR NOTES (2) The MANUAL indicator light will then be Obj. V. lit and all error and alarm indications that were on prior to bypass will be blanked out on the RWM system display (3) A manual bypass will also light the RWM and PROGR indicator on the RWM-COMP-PROGR-BUFF pushbutton.
f. SYSTEM INITIALIZE pushbutton switch/indicator (1) The SYSTEM INITIALIZE switch is
, depressed to initialize the RWM syste (2) Initialization must be performed whenever the RWM has been taken off line, as occurs whenever the RWM program is aborted or manually bypasse (3) Therefore, following any program abort or bypass, the SYSTEM INITIALIZE switch must be depressed before the program can be run agai (4) The SYSTEM INITIALIZE window lights white while the switch is held down.
g. SYSTEM DIAGNOSTIC switch/indicator (1 ) This switch can be pressed at any time after the system has been initialized to request that the system diagnostic routine be performe (2) The RWM program will thereupon be initiated and will perform the routine, which consists of applying and then removing in sequence the insert and withdraw blocks (nominal 10 second frequency).
(3) The operator can verify the operability NOTE: Rod insert of the rod block circuits by observing and withdrawal that the INSERT BLOCK and permit lights will go WITHDRAW BLOCK alarm lights come off when block is on and then go off as the blocks are applie BFN Control Rod Drive System 1-01-85 Unit 1 Rev. 0005 Page 19 of 179 Rod Worth Minimizer (RWM) (continued) For group limits only, RWM recognizes the Nominal Limits only. The Nominal Limit is the insert or withdraw limit for the group assigned by RWM. The Alternate Limit is no longer recognized by the RWM as an Acceptable Group Limi During RWM latching, the latched group will be the highest numbered group with 2 or less insert errors and having at least 1 rod withdrawn past its insert limit . With Sequence Control ON, latching occurs as follows: (Normally, startups will be performed with Sequence Control ON) RWM will latch down when all rods in the presently latched group have been inserted to the group insert limit and a rod in the next lower group is selecte RWM will latch up when a rod within the next higher group is selected, provided that no more than two insert errors resul . With Sequence Control OFF, latching occurs as follows: For non-repeating groups, latching occurs as described above, OR b, For repeating groups, latching occurs to the next setup or set down based on rod movement as opposed to rod selectio Latching occurs at the following times: System initializatio . Following a "System Diagnostic" reques . When operator demands entry or termination of "Rod Test." When power drops below LPA . When power drops below LPS . Every five seconds in the transition zon . Following any full control rod scan when power is below LPA . Upon demand by the Operator (Scan/Latch Request function). Following correction of insert or withdraw error BFN Control Rod Drive System 1-01-85 Unit 1 Rev.OOOS Page 136 of 179 8.18 Reinitialization of the Rod Worth Minimizer
[1 ] VERIFY the following initial conditions are satisfied: * The Rod Worth Minimizer is available to be placed in operation 0 * Integrated Computer System (lCS) is available 0 * The Shift Manager/Reactor Engineer has directed reinitialization of the Rod Worth Minimizer 0 [2] REVIEW all Precautions and Limitations in Section [3] VERIFY RWM SWITCH PANEL, 1-XS-85-9025 in NORMA [4] CHECK the Manual/Auto Bypass lights are extinguishe [5] DEPRESS AND HOLD INOP/RESET pushbutto [6] CHECK all four lights (RWM/COMP/PROG/BUFF) are illuminate D [7] RELEASE INOP/RESET pushbutton and CHECK all four lights extinguishe D [8] SIMUL TANEOUSL Y DEPRESS OUT OF SEQUENCE/SYSTEM INITIALIZE pushbutton and INOP/RESET pushbutton to place the Rod Worth Minimizer in servic D [9] IF Rod Worth Minimizer will NOT initialize, THEN DETERMINE alarms on RWM Display Screen and CORRECT problem [10] IF unable to correct problems and initialize RWM, THEN NOTIFY Reactor Enginee . RO 202001K6.09 00 lICIAlT2G2/68 - RECIRC124120200 lK6.091IRO/SRO/BANK Given the following plant conditions: * Unit 3 is operating at 35% power with 'A' & 'C' Reactor Feed Pumps (RFP)
running and 'B' RFP idlin * Both Recirculation Pump speeds are 53%.
* The 'A' RFP trips, resulting in the following conditions:
REACTOR WATER LEVEL ABNORMAL (2-ARP-9-5A Window 8).
REACTOR CHANNEL 'A' AUTO SCRAM (2-ARP-9-5B Window 1).
REACTOR CHANNEL 'B' AUTO SCRAM (2-ARP-9-5B Window 2).
- Indicated Reactor Water Level drops to (-) 43" before 'B' RFP is brought on line to reverse the level trend and level is stabilized at 33".
Which ONE of the following describes the steady state speed of both Recirculation Pumps? A.II Running at 28% spee Running at 45% spee Running at 53% spee Tripped on ATWS/RPT signa KIA Statement: 202001 Recirculation K6.09 - Knowledge of the effect that a loss or malfunction of the following will have on the RECIRCULATION SYSTEM: Reactor water level KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions and times to determine the effect of a change in reactor water level on the Recirculation Syste References: 3-01-68, OPL 171.007, OPL 171.012 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 outcom NRC Exam Page: lof2 2/20/2008
REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Did plant conditions exceed the Recirc Runback setpoin . Which Runback is appropriate for the given condition A - correct: Total Feed Flow would drop below 19% when the 'A' RFP trips thus impacting reactor water level control and initiating a Recirc Runback to 28%. B - incorrect: This is plausible because a Recirc Runback DID occur, but the 45% speed given in the distractor is the typical speed the Recirc Pumps run at during startup, not following a RFP tri C - incorrect: This is plausible based on the initial power level being close enough to create doubt on Total Feed Flow resulting from the trip of one RF D - incorrect: This is plausible because ATWS/RPT signals are associated with low RPV level, however the setpoint is -45 inches and level only lowered to -43 inche Amplification: This transient has been run on the simulator and can be duplicated if necessar Note - Changed the power level in the Stem from 55% to 35% so that desired response is achievable. Change the water level from -10" to -43" in the 4th bullet for plausibilty.
Page: 20f2 2120/2008
OPL 171.007 Revision 22 Page 26 of 86 The buoyancy of the steam bubbles has a similar effect. The effects can be seen in the natural circulation line, and shows the power versus flow with no recirculation pumps runnin ) As core power rises along either line, the rise in core flow rate eventually stops, and will actually lower if power continues to raise. This is a result of the additional flow resistance created by the additional two-phase flo ) The rises due to density differences and buoyancy are not as pronounced at higher power levels because higher steam production tends to cause a pressure drop across the core, which retards core flow rise. The 100% rod line (or the 100% load line) is based on equilibrium xenon conditions and is defined by the core configuration that will result in rated core thermal power when the core flow is raised to rated flow. Other flow control lines are defined similarly; for example, the 80% rod line will yield 80% of rated core thermal power at rated core flow. The minimum expected flow control line defines the relationship between power and flow when control rods are pulled until 19% feed flow is exceeded. At that point the 28% limiter is removed, and flow could be raised to maximum. The pump constant speed line (Pts G to 0 or H to I) is the rated recirculation pump speed line. The shape of the line will be followed if core power is changed with no change in recirculation pump speed (e.g., control rod movement or end-of-cycle coast-down). Upon a power reduction, a core flow will raise due to a reduction in the fuel channel void fraction and a subsequent reduction in two-phase flow resistanc The reverse is true for a power ris OPL 171.007 Revision 22 Page 27 of 86 The recirculation pump and jet pump NPSH As the pressure limit lines are designed to prevent cavitation drops in the throat of of the jet pumps and the recirculation the jet pump, the pumps. Notice that at rated core flow, temp at which boiling cavitation of the recirculation pumps or the starts will lower. If jet pumps will not occur if feedwater flow is the boiling point above 19% of rate drops to or below the actual temp of the The minimum power line is a flow interlock water, then cavitation line to prevent cavitation of the jet pump will occu The recirculation pumps cannot be operated above 28% pump speed until measured feedwater flow is greater than 19% of rate ) Similarly, the recirculation pump will run back to minimum speed (28%) if the measured feedwater flow becomes less than 19% while the recirculation pump speed is greater than minimu ) This assures adequate subcooling (and therefore adequate NPSH) for all normal modes of pump operation. The Increased Core Flow area is bounded Normally operate at by points D, G, H, and I and allows a 107-109% rod line at maximum flow of 107.62 x 106 Ibm/hr at 100% power (initial 100% power. As power is lowered from target rod pattern) 100% to 40.4%, flow rises from 105% to 112%. This area can be used to lengthen the cycle when all rods are fully withdrawn. The Maximum Extended Load Line region is defined by the area above the 100% Rod Line and below the Maximum Extended Spectral Shift Load Line Limit Analysis (MELLLA) Lin techniques raise fast This region is part of the Normal Operating neutron flux during Region. The MELLLA Line has been early core life. The determined by analysis to be the most fast neutrons react limiting power/flow condition at which the with U238 to form unit will operate at steady state condition PU239. PU239 will The purpose of this line and region is to fission with thermal ensure that operation at the 100% Rod Line neutrons. This is easily achieved at full power and to take extends fuel life and advantage of the neutron spectral shift cycle length by which occurs at higher power/flow lines and "making" fuel during which raises fuel economy. The MELLLA the cycl region is bounded at 113.6% rod line and 81 % core flo BFN Reactor Recirculation System 3-01-68 Unit 3 Rev. 0066 Page 13 of 179 PRECAUTIONS AND LIMITATIONS (continued) 10. The out of service pump may NOT be started unless the temperature of the coolant between the operating and idle Recirc loops are within 50°F of each other. This 50°F delta T limit is based on stress analysis for reactor nozzles, stress analysis for reactor recirculation components and piping, and fuel thermal limits. [GE SIL 517 Supplement 1] 11. The out of service pump may NOT be started unless the reactor is verified outside of regions 1, 2 and 3 of the Unit 3 Power to Flow Map (ICS or Station Reactor Engineering, 0-TI-248).
12. The temperature of the coolant between the dome and the idle Recirc loop should be maintained within 75°F of each other. If this limit cannot be maintained a plant cooldown should be initiated. Failure to maintain this limit and NOT cooldown could result in hangers and/or shock suppressers exceeding their maximum travel range. [GE SIL 251, 430 and 517] Recire Pump controller limits are as follows: When any individual RFP flow is less than 19% and reactor water level is below 27 inches, speed limit is set to 75%(-1130 RPM speed) and if speed is greater than 75%(-1130 RPM speed), Recirc speed will run back to 75%(-1130 RPM speed). When total feed water flow is less than 19% (15 sec TO) or Recire Pump discharge valve is less than 90% open, speed limit is set to 28%
(-480 RPM speed) and if speed is greater than 28%{-480 RPM speed),
Recirc speed will run back to 28%(-480 RPM speed).
BFN Reactor Recirculation System 3-01-68 Unit 3 Rev. 0066 ( Page 15 of 179 PRECAUTIONS AND LIMITATIONS (continued) The power supplies to the MMR and DFR relays are listed belo VFD3A I&C BUS A (BKR 215) 3-RLY-068-MMR3/A & DFR3/A ICS PNL 532 (BKR 30) 3-RL Y-068-MMR2/A & DFR2/A UNIT PFD (BKR 615) 3-RL Y-068-MMR1/A & DFR1/A VFD 3B I&C BUS B (BKR 315) 3-RL Y-068-MMR3/B & DFR3/B ICS PNL 532 (BKR 26) 3-RL Y-068-MMR2/B & DFR2/B UNIT PFD (BKR 616) 3-RLY-068-MMR1/B & DFR1/B A complete list of Recirc System trip functions is provided in Illustration 4. The RPT breakers between the recirc drives and pump motors will open on any of the following: Reactor dome Pressure ~ 1148 psig (ATWS/RPT). (Both pressure switches in Logic A or both pressure switches in Logic B will cause RPT breakers to trip both pumps.) (2 out of 2 taken once logic) Reactor Water Level ~ -45" (ATWS/RPT). (Both level switches in Logic A or both level switches in Level B will cause RPT breakers to trip both pumps.) (2 out of 2 taken once logic) Turbine trip or load reject condition, when ~ 30% power by turbine first stage pressure (EOC/RPT). The ATWS/RPT A(B) logic to trip the RPT breakers is defeated if the ATWS/RPT/ARI A(B) manual logic is armed using the arming collar on Panel 3-9-5. B(A) logic would still be functional and trip the RPT breakers if the setpoints are reached. If both manual push-buttons on 3-9-5 are armed, A TWS/RPT automatic logic is totally defeated (no RPT breaker trip will occur if the ATWS/RPT trip setpoints are reached). EOC/RPT logic and ATWS/ARI logic will function without regard to the position of the arming collar ATWS/RPT/ARI logic can be reset 30 seconds after setpoints are rese NRC RO EXAM 30. RO 215001Al.01 001IMEM/T2G2/TIP!I215001Al.OllIRO/SROINEW 10116107 Given the following Unit 2 conditions:
* Preparations are in progress to commence a startup following a refueling outag * The Drywell Close-out inspection is in progress with three personnel inside the Drywel * Reactor Engineers have begun running Traversing In-Core Probe (TIP) trace * RX BLDG AREA RADIATION HIGH (2-XA-9-3A, Window 22) is in alar * Area Radiation Monitor 2-RA-90-1 D indicates 16 mr/h Which ONE of the following describes the required actions?
A." This is an expected indication while running TIP traces, therefore TIP operation may continu Verify the TIP mechanism automatically shifts to REVERSE mode and begins a full retractio Manually retract the detector in FAST speed to the In-Shield position and close the ball valv Evacuate the Unit-2 Drywell and Reactor Building. Allow the TIP trace to run automatically to completio KIA Statement: 215001 Traversing In-core Probe A 1.01 - Ability to predict and/or monitor changes in parameters associated with operating the TRAVERSING IN-CORE PROBE controls including: Radiation levels:
(Not-BWR1)
KIA Justification: This question satisfies the K/A statement by requiring the candidate to determine the operating limitations of the TIP system with respect to high radiatio References: 2-01-94 Precautions & Limitations Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam Friday, February 29, 2008 3:01 :05 AM 65
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: Running TIP traces will cause radiation levels inside the Reactor Building and the Drywell to increase due to the high activity associated with the TIP detectors as they are withdrawn from their shield and inserted into the appropriate TIP Tube which penetrates the Drywell and Reactor Vessel. As a result, limitations are provided to ensure personnel are protected from unnecessary exposure during performance of TIP traces. Although the given radiation alarms and conditions are typical during operation of the TIP System, with personnel inside the Drywell there are special considerations which must be addressed because of these radiation conditions which are a DIRECT result of TIP System operation. In order to correctly answer this question the candidate must recognize that high radiation conditions will exist inside the Drywell, but will be limited to areas inside the Reactor Vessel pedestal which is not typically entered during a close-out inspection. In addition, TIP operation with personnel inside the Drywell is permissable with appropriate approval and notification of personnel in the Drywel C - correct: A - incorrect: This is plausible because that limitation is placed on TIP operation, but ONLY when TIP operation is NO longer required. The TIP detector can be stored in the Indexer in-between traces using the same TIP Machine for ALARA concern B - incorrect: This is plausible because specific permission and controls are required to allow this condition, but it is allowabl D - incorrect: This is plausible because the TIP response to a PCIS isolation is correct, but it is not a Group 6 isolation.
Friday, February 29, 2008 3:01 :05 AM 66
BFN Traversing Incore Probe System 2-01-94 Unit 2 Rev. 0029 Page 7 of 26 PRECAUTIONS AND LIMITATIONS [NER/C] Verification of a digit in CORE LIMIT and DETECTOR POSITION windows prior to or during TIP insertion ensures TIPs retain the ability to determine its proper position. This will prevent malfunctions which could damage the TIP detector. [GE SIL-166] To prevent accidental exposure to personnel, immediately evacuate the area if the TIP drive area radiation monitor alarm [NER/C] Always observe READY light illuminated prior to inserting detecto [GE SIL-166] [NER/C] DO NOT move CHANNEL SELECT switch with detector inserted past Indexer position (0001). The common channel interlock can be defeated in this manner resulting in detector and equipment damage. [GE SIL-092] [NER/C] Should detector fail to shift to slow speed when it enters the core, the LOW switch should be turned on, switched to manual mode, and the detector withdrawn. [GE SIL-166] [NER/C] Length of time detector is left in core should be minimized to limit activation of detector and cable. [GE SIL-166] [NER/C] When TIP System operation is not desired, detectors should be retracted and stored in chamber shield with ball valves closed. [GE SIL-166] Storage of detector in Indexer (0001) is allowed only for ALARA concerns and to prevent unnecessary masking of multiple inputs to annunciator RX BLDG AREA RADIATION HIGH 2-RA-90-1 D (2-XA-55-3A, Window 22). [NER/C] Upon receipt of a PCISsignal (low reactor water level or high drywell pressure), any detector inserted beyond its shield chamber should be verified to automatically shift to reverse mode and begin withdrawal. Once in shield, ball and purge valves close. [GE SIL-166] Ball valve cannot be reopened until PCIS is reset on Panel 2-9-4 and manual reset of TIP ISOLATION RESET pushbutton 2-HS-94-7D/S2 located on Panel 2-9-1 A detector should not be abruptly stopped from fast speed to off without first switching to slow spee [NER/C] Drive Control Units (DCU) should be monitored during withdrawal to prevent any chamber shield withdrawal limit from being overrun. Detectors should be stopped manually at shield limit if auto stop limit switch should fail and verify ball valve closes. [GE SIL-166] Only one TIP at a time should be operated when maintenance is being performed in TI P drive are BFN Traversing Incore Probe System '2-01-94 Unit 2 Rev. 0029 Page 8 of 26 PRECAUTIONS AND LIMITATIONS (continued) [NRC/C] DO NOT operate TIPswith personnel inside TIP Room or in vicinity of TIP tubing and Indexers in Drywell. Requirement may be waived with approval of Shift Manager and site RADCON manager or designee. In this instance, RADCON is required to establish such controls as are necessary to prevent access to TIP tubing and Indexer areas to preclude unnecessary exposure to personnel working in Drywell. RADCON Field Operations Shift Supervisor is required to be notified prior to operation of TIP System. [NRC Information Notice 88-063, Supplement 2] No channel should be indexed to common channel 10 unless all other channels are not indexed to channel 10 and all their READY lights are illuminate [NER/C] DO NOT turn MODE switch to OFF on Drive Control Unit if detector is outside shield chamber unless personnel safety requires it. [GE SIL-166] This removes power preventing automatic withdrawal on PCIS signal and causing ball valves to close on cable or detector. Tip Ball Valves CANNOT fully close and shear valves may have to be actuate CHANNEL SELECT switches on Drive Control Units should always be rotated in clockwise direction when selecting channel Connector on shear valve indicator circuit should not be removed while testing shear valve explosive charges or performing shear valve maintenance with detector inserted. This will cause an automatic detector withdrawa Continuous voice communication should be maintained between TIP operator or maintenance personnel in control room and drive mechanism area while maintenance is being performed and TIP detector driving is necessar Each applicable ball valve should be opened prior to operating that TIP machin TIP Drive Mechanisms and Indexers should have continuous purge supply unless required to be removed from service for maintenanc During outages when containment is deinerted for personnel access, TIP Indexer purge supply should be transferred from nitrogen to Control Air for personnel safet Detector damage is possible if TI P ball valve is left open, or is opened during DRYWELL PRESSURE TEST. (GE SIL-166)
0610 NRC RO EXAM 31 . RO 216000Kl.1 0 00 lIMEMiT2G2/PR.INSTRl9/216000Kl.1 01IRO/SROIBANK Which ONE of the following indicates how raising recirculation flow affects the Emergency System (Wide Range) 3-58A/58B and Normal Control Range Level Indicators (e.g., L1-3-53) on Panel 9-5? Emergency System Range indication will _ _ _ _ and the Normal Control Range indication will - - - - - indicate lower; indicate lowe NOT be affected; indicate highe indicate higher; NOT be affecte .01 indicate lower; NOT be affecte KIA Statement: 216000 Nuclear Boiler Inst K1.10 - Knowledge of the physical connections and/or cause- effect relationships between NUCLEAR BOILER INSTRUMENTATION and the following: Recirculation flow control system KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific knowledge of the effect of changes in Recirculation flow on reactor water level instrumentatio References: OPL 171.003 Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam Friday, February 29, 20083:01 :05 AM 67
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: Normal Control Range level instruments are calibrated under hot conditions with Recirc pumps running. Based on that and the location of the variable leg taps of these instruments, changes in Recirc flow do not affect their indication. Emergency Systems Range level instruments are claibrated under cold conditions with no Recirc pumps running. As Recirc flow changes; the pressure sensed by the variable leg also changes, thereby causing a change in indicated water level. The magnitude and direction of the change in indication depends on perspective. Specifically, the initial and final conditions used to quantify the chang D - correct: A - incorrect: This is plausible because Emergency System Range instruments will read lower, but the Narrow Range instruments will no B - incorrect: This is plausible because Narrow Range instruments may read slightly higher at colder conditions, but this does NOT apply to Recirc flow change C - incorrect: This is plausible because Narrow Range instruments are not effected by Recirc Flow changes, but Emergency System Range instruments will read lower.
Friday, February 29,20083:01 :05 AM 68
OPL 171.003 Revision 17 Page 20 of 54 INSTRUCTOR NOTES d. Four ranges of level indication (1 ) Normal Control Range (Narrow Range) Obj. V. (a) oto +60 inch range covering the Obj. V. normal operating range (analog) with TP-3 shows only
+60" up to +70" digital and 0" down to analog scale - 10" digital reading (b) Referenced to instrument zero (c) Four of these instruments are used by Feedwater Level Control System (FWLCS). The level signal utilized by the FWLCS is not directed through the Analog Trip Syste Temperature Obj. V.B.1 compensated by a Obj. V.B.1 pressure signal i Most accurate level indication available to the operator ii Calibrated for normal operating pressure and temperature (d) These indicators and a recorder point (average of the four) are located on Panel 9- NOTE: An air bubble or leak in LER 85-006-02 the reference leg can cause (See LP Folder)
inaccurate readings in a non- (Section X.C. conservative direction resulting in provides more a mismatch between level dEltail) indicator This problem is particularly prevalent after extended outages when starting up from cold shutdown conditions and at low reactor pressure OPL171.003 Revision 17 Page 21 of 54 INSTRUCTOR NOTES (e) Four other narrow range Associated with instruments are located in the RFPT/Main Turbine control room, two above the and HPCIIRCIC trip FWLCS level indicators on panel instruments 9-5 (3-20SA & D), one above HPCI (3-20SB)and one above RCIC (3-20SC)on panel 9-3.
(2) Emergency Systems Range (Wide Range) 2 Analog meters and 2 Digital meter (a) -155 to +60 inches range covering normal operating range and down to the lower instrument nozzle return (b) Referenced to instrument zero (c) Four MCR indicators on Panel 9-5 monitor this range of level indicatio (d) Calibrated for normal operating pressure and temperature (e) The level signal utilized by the Wide Range instruments have safety related functions and are directed through the Analog Trip Syste (f) Level indication for this range is Obj. V.B.1 also provided on the Backup Control Panel (25-32).
(3) Shutdown Vessel Flood Range (Flood-up Range)
(a) o to +400 inches range covering upper portion of reactor vessel (b) Referenced to instrument zero Calibrated for cold conditions <<212°F, 0 psig) (c) Provides level indication during vessel flooding or cool dow OPL 171.003 Revision 17 . Page 32 of 54 INSTRUCTOR NOTES Transient flashing effects can cause indicated level to oscillate or be erratic. As the Teference leg refills, the indicated level approaches a more accurate water level indicatio The RVLlS mod decreases the time necessary for this refill to occur j. Normal Control Range (Narrow Range) and Emergency Systems Range (Wide Range) Level Discrepancies (1) Narrow Range level instrumentation is calibrated to be most accurate at rated temperature and pressure (particularly the instruments for FWLCS, since they are temperature compensated). At cold conditions the non-FWLCS instruments read high (not temperature compensated).
(2) Wide Range instruments are also calibrated for rated temperature and pressure (a) The indicated level on the Wide Obj. V.B.15 Range (9-5) is also affected by changes iri the subcooling of recirculation water and the amount of flow at the lower (variable leg) ta (b) At rated conditions with minimum recirculation flow the Wide Range instruments are accurate. As recirculation flow is increased past the lower tap it has a significant velocity head and some friction loss which reduces the pressure on the variable leg to the differential pressure instrument, resulting in an indicated level lower than actual. This could be as much as 10-15 inches error when at rated flow and powe (c) Due to calibration for rated conditions and no density compensation at cold conditions these instruments read hig NRC RO EXAM 32. RO 219000K2.02 OOllCIA/T2G2/0I-74//219000K2.02//RO/SROIBANK Given the following plant conditions:
* Unit 2 is at 100% rated power with Residual Heat Removal (RHR) Loop II in Suppression Pool Cooling mode to support a High Pressure Coolant Injection (HPCI) Full Flow Test surveillanc * Unit 1 experiences a LOCA which results in a Common Accident Signal (CAS) initiation on Unit Which ONE of the following describes the current status of Unit 2 RHR system and what actions must be taken to restore Suppression Pool Cooling on Unit 2?
A. ALL four RHR pumps receive a trip signal. Place RHR Loop II in Suppression Pool Cooling IMMEOIATEL . '2A' and '2C' RHR Pumps are tripped. '28' and '20' pumps are unaffected. NO additional action is require C~ ALL four RHR pumps receive a trip signal. Place RHR Loop II in Suppression Pool Cooling after a 60-second time dela O. '28' and '20' RHR Pumps are tripped. '2A' and '2C' pumps are unaffected. Place RHR Loop I in Suppression Pool Cooling IMMEOIATEL KIA Statement: 219000 RHR/LPCI: Torus/Pool Cooling Mode K2.02 - Knowledge of electrical power supplies to the following: Pumps KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions and times to determine which RHR pumps can be used for Suppression Pool Coolin References: 2-01-74, OPL 171.044 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 outcom NRC Exam Friday, February 29, 2008 3:01 :05 AM 69
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Response of Unit 2 RHR pumps due to a Unit 1 CAS initiatio . Recognize the difference between a Single Unit CAS and Simultaneous Unit CA . Recognize that Preferred and Non-preferred Emergency Core Cooling System (ECCS) Pumps do NOT apply with the given condition Unit 1 Preferred RHR pumps are 1A and 1C. Unit 2 Preferred RHR pumps are 28 and 2D. LOCA signals are divided into two separate signals, one referred to as a Pre Accident Signal (PAS) and the other referred to as a Common Accident Signal (CAS). If a unit receives a CAS, then all its respective RHR and Core Spray pumps will sequence on based upon power source to the SO Boards. All RHR and Core Spray pumps on the non-affected unit will trip (if running) and will be blocked from manual starting for 60 seconds. After 60 seconds all RHR pumps on the non-affected unit may be manually started. The non-preferred pumps on the non-affected unit are also prevented from automatically starting until the affected unit's accident signal is clea The preferred pumps on the non-affected unit are locked out from automatically starting until the affected unit accident signal is clear OR the non-affected unit receives an accident signa C - correct: A - incorrect: This is plausible because all four RHR pumps on Unit 2 will trip, but they are locked out from manual start for 60 seconds based on Diesel Generator and/or Shutdown Board loading concern incorrect: This is plausible based on RHR Loop II being the preferred pumps for Unit D - incorrect: This is plausible if taken from the perspective of Unit 1 operation, NOT Unit 2 operation.
Friday, February 29, 2008 3:01 :05 AM 70
BFN Residual Heat Removal System 2-01-74 Unit 2 Rev. 0133 Page 331 of 367
. Appendix A (Page 2 of 7)
Unit 1 & 2 Core Spray/RHR Logic Discussion ECCS Preferred Pump Logic Concurrent Accident Signals On Unit 1 and Unit 2 With normal power available, the starting and running of RHR pumps on a 4KV Shutdown Board already loaded by the opposite unit's Core Spray, RHR pumps, and RHRSW pumps could overload the affected 4KV Shutdown Boards and trip the normal feeder breaker. This would result in a temporary loss of power to the affected 4KV Shutdown Boards while the boards are being transferred to their diesels. To prevent this undesirable transient, Unit 2 RHR Pumps 2A and 2C are load shed on a Unit 1 accident signal and Unit 1 Pumps 1Band 10 will be load shed on a Unit 2 accident signal. Unit 2 Core Spray Pumps 2A and 2C are load shed on a Unit 1 accident signal and Unit 1 Core Spray Pumps 1Band 10 will be load shed on a Unit 2 accident signal. This makes the Preferred ECCS pumps Unit 1 Division I Core Spray and RHR Pumps and Unit 2 Division 2 Core Spray and RHR Pump Conversely, the Non-preferred ECCS pumps are Unit 1 Division 2 Core Spray and RHR Pumps and Unit 2 Division 1 Core Spray and RHR Pump The preferred and non-preferred ECCS pumps are as follows: UNIT 1 & 2 PREFERRED ECCS Pumps CS1A,CS1C,RHR1A,RHR1C CS 2B, CS 20, RHR 2B, RHR 20 NON-PREFERRED ECCS Pumps CS 1B, CS 10, RHR 1B, RHR1D CS 2A, CS 2C, RHR 2A, RHR2C UNIT3 Unit 3 does not have ECCS Preferred/Non-Preferred Pump Logi Accident Signal On One Unit With an accident on one unit, ECCS Preferred pump logic trips all running RHR and Core Spray pumps on the non-accident uni OPL 171.044 Revision 15 Page 50 of 159 INSTRUCTOR NOTES Note: Presently Unit 1 Accident signal will not affect Unit 2 due to OCN H2735A that lifted wires from relays. Unit 2 will still affect Unit 1. However, the following represents modifications to the inter-tie logic as it will be upon Unit 1 recover Ob VB.1 Ob V. (1) Unit 1 Preferred RHR pumps are 1A and 1C Ob V. Ob V (2) Unit 2 Preferred RHR pumps are 28 and 20 Ob VE.II (3) Unit 2 initiation logic is as follows:Oiv 1 RHR logic initiates Oiv 1 pumps ( A and C), and Oiv 2 logic initiates Oiv 2 pumps (B and D) Accident Signal (1) LOCA signals are divided into two separate Ob V.B.1 signals, one referred to as a Pre Accident Ob V Signal (PAS) and the other referred to as a Ob V. Common Accident Signal (CAS). Ob V Ob V.E.II
* PAS-122" Rx water level (Level 1) Note:
It should be clear OR that the only 2.45 psig OW pressure difference
* CAS between the two signals is the-122" Rx water level (Level 1) inclusion of Rx OR pressure in the CAS signal. The 2.45 psig OW pressure AND <450 PAS signal is an psig Rx pressure anticipatory signal that allows the DG's to start on (2) If a unit receives an accident signal, then all rising OW its respective RHR and Core Spray pumps pressure and be will sequence on based upon power source to ready should a the SO Board CAS be receive (3) All RHR and Core Spray pumps on the non-affected unit will trip (if running) and will be blocked from manual starting for 60 seconds.
(
OPL 171.044 Revision 15 Page 51 of 159 INSTRUCTOR NOTES (4) After 60 seconds all RHR pumps on the non- Operator diligence affected unit may be manually starte required to prevent (5) The non-preferred pumps on the non-overloading SD affected unit are also prevented from boards/DG's automatically starting until the affected unit's accident signal is clea (6) The preferred pumps on the non-affected unit are locked out from automatically starting until the affected unit accident signal is clear OR the non-affected unit receives an accident signal.
g. 4KV Shutdown Board Load Shed Obj. V. (1) A stripping of motor loads on the 4KV boards occurs when the board experiences an undervoltage condition. This is referred to as a 4KV Load Shed. This shed prepares the board for the DG ensuring the DG will tie on to the bus unloaded and without fault (2) The Load Shed occurs when an undervoltage is experienced on the board i.e. or if the Diesel were tied to the board (only source) and one of the units experienced an accident signal which trips the Diesel output breake (3) Then, when the Diesel output breaker interlocks are satisfied, the DG output breaker would close and, if an initiation signal is present (CAS) the RHR, CS, and RHRSW pumps would sequence on (4) Following an initiation of a Common Accident Signal (which trips the diesel breaker), if a subsequent accident signal is received from another unit, a second diesel breaker trip on a
"unit priority" basis is provided to ensure that the Shutdown boards are stripped prior to starting the RHR pumps and other ECCS loads (5) When an accident signal trip of the diesel Occurs due to breakers is initiated from one unit (CASA or actuation of the CASB), subsequent CAS trips of all eight diesel breaker diesel breakers are blocke TSCRN relay
0610 NRC RO EXAM 33. RO 226001A4.12 OOl/MEMlT2G2/PC/P11226001A4.12/3.8/3.9IRO/SROIBANK A LOCA inside containment results in the following plant conditions:
* Drywell pressure is 20 psig * Drywell temperature is 210°F * Suppression chamber pressure is 18 psig * Suppression chamber temperature is 155°F * Suppression pool level is (+) 2 inches * Reactor water level is (+) 30 inches Which ONE list of parameters below must ALWAYS be addressed to determine when it is appropriate to spray the drywell? Suppression Chamber temperature - Drywell pressure - Drywell temperature Suppression Chamber pressure - Drywell temperature - Suppression Pool level C." - Drywell pressure - Drywell temperature - Reactor water level Reactor water level - Suppression Chamber temperature - Drywell pressure Friday, February 29, 2008 3:01 :05 AM 71
0610 NRC RO EXAM KIA Statement: 226001 RHR/LPCI: CTMT Spray Mode A4.12 - Ability to manually operate and/or monitor in the control room: Containmentldrywell pressure KIA Justification: This question satisfies the K/A statement by requiring the candidate to use specific knowledge of which containment parameters are used to determine when Containment Sprays can be use References: 1/2/3-EOI-2 Flowchart Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Orywell temperature and pressure are always required to ensure Curve 5 limits are not exceede . RPV level is always required to verify adequate core cooling is assured prior to diverting RHR flow for Orywell spray . Suppression Pool level is always required to verify Suppression Chamber to Orywell vacuum breakers are uncovere . Suppression Chamber pressure is ONLY required when initiating Orywell Sprays from flowpath PC/ . Suppression Chamber temperature is NOT required to initiate Orywell Spray C - correct: A - incorrect: This is plausible because OW temp and press are required, but Suppression Chamber (Torus) temp is NO B - incorrect: This is plausible because OW temp and SP level are required, but SC press is ONLY required when initiating OW Sprays using PC/Po D - incorrect: This is plausible because RPV level and OW press are required, but Suppression Chamber (Torus) temp is NOT.
Friday, February 29, 2008 3:01 :05 AM 72
\"iHEN SUP!'>R CHM5R PRESS EXCEEDS ~2 PSI THEN CONTINUE IN THIS PROCEDURE L
?CtP-!j IS SU?PR Pt.!.. Vl
~IO L BELOW lB FT ARE DW1EMPAND NO L OW PRESS WITHIN THE SAFE ARE.t.. OF CURVE 5 ?CiP-l YES L SHUT DOWN RECIRC PUMPS .A.\JD DW BLOWERS L
CAUTION PUMP NPSH AND VORTEX LIMITS L 1H1TIAT6 rJ;N SPRAYS-USING QblLYPUMPS,WIREQUiRED TO ASSUREADEQUATE GORE COOlING 'BY CONTINUOUS 1t4J{APPX HB} L
I! +i L L ,i IMT.Il1.T£ 0'1. S'AAVS USM Oi>LYFU'~ M:)! RecuF/$o tO/.~L.'!'l1S ~fJA'EO~CCiiJ~3 II!w ~fJ!i
'NJ iJoJi"PX17e1 L
0610 NRC RO EXAM 34. RO 234000G2.4.50 001!C/A/T2G2///234000G2.4.50//RO/SROINEW 10/16107 Given the following plant conditions:
* Fuel movement is in progress for channel changeout activities in the Fuel Prep Machin * Gas bubbles are visible coming from the de-channeled bundl * An Area Radiation Monitor adjacent to the Spent Fuel Storage Pool begins alarmin Which ONE of the following describes the action (s) to take?
Immediately STOP fuel handling, then _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ evacuate ALL personnel from the Refuel Floo notify RADCON to monitor & evaluate radiation level C." evacuate non-essential personnel from the Refuel Floo obtain Reactor Engineering Supervisor's recommendation for movement and sipping of the damaged fuel assembl KIA Statement: 234000 Fuel Handling Equipment 2.4.50 - Emergency Procedures I Plan Ability to verify system alarm setpoints and operate controls identified in the alarm response manual KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the corrective actions involving Fuel Handling equipment under emergency condition References: 1/2/3-AOI-79-1 & 79-2, 1/2/3-ARP-9-3A (W1) 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 outcom NRC Exam Friday, February 29, 2008 3:01 :05 AM 73
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Whether indications are consistent with fuel damage or inadvertant criticalit . Based on the answer to Item #1 above, enter the appropriate AO . Immediate Operator Actions for the selected procedure, AOI-79-1, Fuel Damage During Refuelin C - correct: Non-essential personnel evacuation is an IMMEDIATE action in AOI-79-1, Fuel Damage During Refuelin A - incorrect: This is plausible because evacuation of ALL personnel is an IMMEDIATE action in AOI-79-2, Inadvertent Criticality During In-Core Fuel Movement However, non-essential personnel evacuation is an IMMEDIATE action in the AOI-79-1 B - incorrect: This is plausible because RADCON notification is a subsequent action in AOI-79-1. However, non-essential personnel evacuation is an IMMEDIATE actio D - incorrect: This is plausible because Reactor Engineer recommendations are a subsequent action in AOI-79-1. However, non-essential personnel evacuation is an IMMEDIATE action.
Friday, February 29,20083:01 :05 AM 74
BFN Panel 9-3 2-ARP-9-3A Unit 2 2-XA-55-3A Rev. 0036 Page 4 of 50 Sensor/Trip Point: FUEL POOL FLOOR AREA RADIATION HIGH RI-90-1 B RI-90-2B For setpoints 2-RA-90-1A RI-90-3B REFER TO 2-SIMI-90 (Page 1 of 1) Sensor RE-90-1 B EI664' R-11 P-LiNE Location: RE-90-2B E1664' R-10 U-LlNE RE-90-3B EI639' . R-10Q-LlNE Probable A. Change in general radiation levels.
Cause: B. Refueling acciden C. Sensor malfunction.
Automatic None Action: Operator A. CHECK 2-RI-90-1A, 2-RI-90-2A and 2-RI-90-3A on Panel 2-9-1 o Action: B. NOTIFY refuel floor personne o C. IF Dry Cask loading/unloading activities are in progress, THEN NOTIFY Cask Superviso .0 D. IF airborne levels rise by 100 DAC AND RADCON confirms, THEN REFER TO EPIP- o E. REFER TO 2-AOI-79-1 or 2-AOI-79-2 as applicabl o F. IF this alarm is not valid, THEN REFER TO 0-01-5 o G. IF this alarm is valid, THEN MONITOR the other parameters that input to it frequently. These other parameters will be masked from alarming while this alarm is sealed i o H. ENTER 2-EOI-3 Flowchar o References: 0-47E600-13 2-47E61 0-90-1 2-45E620-3 GE 730E356 Series, TVA Calc NDQ00902005001/EDC63693
BFN Fuel Damage During Refueling 2-AOI-79-1 Unit 2 Rev. 0017 Page 3 of 7 PURPOSE This instruction provides the symptoms, automatic actions and operator actions for a fuel damage accident. SYMPTOMS Possible annunciators in alarm: FUEL POOL FLOOR AREA RADIATION HIGH (2-XA-55-3A, window 1). AIR PARTICULATE MONITOR RADIATION HIGH (2-XA-55-3A, window 2). RX BLDG, TURB BLDG, RF ZONE EXH RADIATION HIGH (2-XA-55-3A, window 4). REACTOR ZONE EXHAUST RADIATION HIGH (2-XA-55-3A, window 21). RX BLDG AREA RADIATION HIGH (2-XA-55-3A, window 22). REFUELING ZONE EXHAUST RADIATION HIGH (2-XA-55-3A, window 34). Gas bubbles visible, in the Spent Fuel Storage Pool and/or Reactor Cavity, attributed to physical fuel damag Known dropped or physically damaged fuel bundl Portable CAM in alar Radiation level on the Refuel Floor is greater than 25 mr/hr and cause is unknow BFN Fuel Damage During Refueling 2-AOI-79-1 Unit 2 Rev. 0017 Page 5 of 7 OPERATOR ACTIONS Immediate Actions
[1] STOP all fuel handlin o [2] EVACUATE all non-essential personnel from Refuel Floo o Subsequent Actions CAUTION The release of iodine is of major concern. If gas bubbles are identified at any time, Iodine release should be assumed until RADCON determines otherwis [1 ] VERIFY secondary containment is intac (REFER TO Tech Spec 3.6.4.1) o [2] IF any EOI entry condition is met, THEN ENTER the appropriate EOI(s). o [3] VERIFY automatic action o [4] NOTIFY RADCON to perform the following: * EVALUATE the radiation level * MAKE recommendation for personnel acces * MONITOR around the Reactor Building Equipment Hatch, at levels below the Refuel Floor, for possible spread of the releas [5] REFER TO EPIP-1 for proper notificatio o
BFN Fuel Damage During Refueling 2-AOI-79-1 Unit 2 Rev. 0017 Page 6 of 7 Subsequent Actions (continued)
[6] MONITOR radiation levels, for the affected areas, using the following radiation recorders and indicators: RR-90-1 (points 1 and 2), 2-MON-90-50 (Address 11),
2-RR-90-142 and 2-RR-90-140 (Panel 2-9-2). 0 RM-90-142, 2-RM-90-140, 2-RM-90-143 and 2-RM-90-141 Detectors A and B (Panel 2-9.c10). 0 RI-90-1A and 2-RI-90-2A (Panel 2-9-11). O* CONS-90-362A (Address 09,10,08) for Unit 1,2, 3-RM-90-250, respectively (Panel 1-9-44). 0
[7] IF possible, MONITOR portable CAMs & ARM [8] REQUEST Chemistry to perform 0-SI-4.8.B.2-1 to determine if iodine concentration has rise [9] NOTIFY Reactor Engineering Supervisor, or his designee, and OBTAIN recommendation for movement and sipping of the damaged fuel assembl [10] OBTAIN Plant Managers approval prior to resuming any fuel transfer operation [11 ] WHEN condition has cleared AND if required, THEN RETURN ventilation systems, including SGTS, to norma REFER TO 2-01-30A, 2-01-30B, 0-01-30F, 0-01-31, and 0-01-6 BFN Inadvertent Criticality During Incore 2-AOI-79-2 Unit 2 Fuel Movements Rev. 0013 Page 5 of 8 OPERA TOR ACTIONS Immediate Actions [1 ] IF unexpected criticality is observed following control rod withdrawal, THEN REINSERT the control ro [2] IF all control rods CANNOT be fully inserted, THEN MANUALLY SCRAM the reacto [3] IF unexpected criticality is observed following the insertion of a fuel assembly, THEN PERFORM the following: 0 [3.1 ] VERIFY fuel grapple latched onto the fuel assembly handle AND immediately REMOVE the fuel assembly from the reactor cor [3.2] IF the reactor can be determined to be subcritical AND no radiological hazard is apparent, THEN PLACE the fuel assembly in a spent fuel storage pool location with the least possible number of surrounding fuel assemblies, leaving the fuel grapple latched to the fuel assembly handl [3.3] IF the reactor CANNOT be determined to be subcritical OR adverse radiological conditions exist, THEN TRAVERSE the refueling bridge and fuel assembly away from the reactor core,preferably to the area of the cattle chute, AND CONTINUE at Step 4.1 [4]. 0 [4] IF the reactor CANNOT be determined to be subcritical OR adverse radiological conditions exist, THEN EVACUATE the refuel floo NRC RO EXAM 35. RO 245000K6.04 001lC/A/T2G2/0I-3511245000K6.041IRO/SROINEW 11128/07 RMS Given the following plant conditions:
(
* Unit 2 is operating at 100% powe * Main Generator is at 1150 MW * The Chattanooga Load Coordinator requires a 0.95 lagging power facto * Generator hydrogen pressure is 65 psi Which ONE of the following describes the required action and reason if Generator hydrogen pressure drops to 45 psig?
Reduce ______________________________________________________ REFERENCE PROVIDED A. generator load below 800 MWe. Pole slippage will NOT occur at this generator loa B~ generator load below 800 MWe. Sufficient cooling capability still exists at this hydrogen pressur C. excitation to obtain a power factor of unity to maintain current generator load. Pole slippage will NOT occur at this power facto D. excitation to obtain a power factor of unity to maintain current generator loa Sufficient cooling capability still exists at this hydrogen pressur Friday, February 29,20083:01 :05 AM 75
0610 NRC RO EXAM KIA Statement: 245000 Main Turbine Gen. I Au K6.04 - Knowledge of the effect that a loss or malfunction of the following will have on the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS: Hydrogen cooling KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the effect of a loss of hydrogen cooling on Main Generator operatio Reference Provided: Generator Capability Curve without axis labeled 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 outcom NRC Exam Friday, February 29, 2008 3:01 :05 AM 76
0610 NRC RO EXAM REFERENCE PROVIDED: Generator Capability Curve without the axis labele Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Current operating point on the Generator Capability Curve based on given condition . Recognize that pole slippage is only a concern when operating with a significant leading power facto . Recognize that pole slippage is a result of under excitation, not excessive generator loa . Recognize that generator hydrogen pressure is directly related to cooling capabilit B - correct: A - incorrect: This is plausible because generator load is properly reduced; but, the basis for the reduction is NOT related to slipping pole C - incorrect: This is plausible because reducing excitation DOES reduce heat generation within the generator; but, NOT sufficiently to prevent generator damage. However, pole slippage is NOTa concern at a unity power facto D - incorrect: This is plausible because reducing excitation DOES reduce heat generation within the generator; but NOT sufficiently to prevent generator damage. In addition, insufficient hydrogen pressure exists at the current generator load even wih a power factor of unity.
Friday, February 29,20083:01 :05 AM 77
BFN Turbine-Generator System 2-01-47 Unit 2 Rev. 0138 Page 214 of 223 Illustration 6 (Page 1 of 1) Generator Kilovar Limitations (Capability Curve) NOTES 1) Operation in this area allowed only during calibration or testing.
2) DO NOT exceed the limits of the capability curve.
3) Maximum LAGGING MVARs are as follows. These values are based on turbine and generator load limitations. More limiting values based on transmission load limits and operational concerns can be found in the Switchyard Manual, 0-GOI-300-4... r=
* 400 MVAR per unit for 1-Unit Operation * 300 MVAR per unit for 2-Unit Operation * 270 MVAR per unit for 3-Unit Operation 4) Curve AB is limited by FIELD heating.
5) Curve BC is limited by ARMATURE heating.
6) Curve CD is limited by ARMATURE CORE END heatin ESTIMATED REACTIVE CAPABILITY CURVES
-'+ Polo 1800 RPM 1330000 kVA 22000 Volts '0.950 PF 0.580 SCR 75.00 PSIG H2 Pressure 510 Volts EX{)itation 46 Oeg. C Cold Gas 617 ft AJritude
,,
,
,
, I I
,,
I I
---------~~------ .,
I **** ~_A I
, ____ ~ __ _ I
...', *1*, -- -----.-
cDC 400 2eo
,
---__ ~--------*--,------~--------~------~*-+----~~-41~
,
._A * * _ _ _ ~~1 ____ * __ .',.~ .,_....._, ...,.,.,.
__ ......, __,_
-_
T-.------- M0~.-W---~- __ ---
'i 030 : \
J
,
I
- 8CO
~_0 _____ ~-._.~"0 __ ___
.~ W __ o/-.*~.--- ____ + _____ ** __ * ____ _ _
,,
" j o 2C\l 400 161)(j MEGAWATTS
I TI EFE *NCE IDE T I TE
ESTIMATED REACTIVE CAPABIUTY CURVES 4 Pola 1800 RPM 1330000 kVA 22000 Volts 0.950 PF 0.580 SCR 15.00 PS!G H2 Pressure 510 vons EX(:ltation 46 Dog. C Cold Gas 617 Ft Attitude
,
.i ' ..
,
j Base Capfibility I , I ,
,
I
, ,
I I HIOO -----~---------~-----**--t--------*t-------- , . -----~---------i---------
.
I
,
,
I I , f I I f i 800 60(; 400 2CO
'"" N '" _""'" ., ~~ _ _ _ * _ _ ",-."
vr:uo
, ,
,
* - ".,.,.,.*.." .... _ _ _ _ ... .., ~ .,~ ,.. Vq v
,
>>_~ ____
.
~~~.,..*t I
I
, ________ * _____ ~_~M4-------
,
,r-~~--.--.--_4--,~------~
,
I I I , DS803~2a
,
I
* .j I I I
I . j" * ( {) 2C(! 14CO
0610 NRC RO EXAM 3 RO 268000A2.01 002/MEM/T2G2111268000A2.01llRO/SROINEW 2/19/08 Given the following plant conditions: (
* BFN is in the process of discharging the Waste Sample Tank to the Unit 1 Condensate Storage Tan * While moving resin containers, a forklift operator accidently punctures the discharge line 4 feet downstream of 0-RR-90-130 (Radwaste Effluent Radiation Monitor).
- Water immediately begins spraying out of the ruptur Which ONE of the following describes the expected system response?
The discharge will _ _ _--->.(....:..1L-)_ _ _ _ because the _ _ _ _..>.:(2=...<)_ _ _ __ automatically terminate without any additional operator action; B. Radwaste Discharge Isolation Valve will auto close on HIGH
- dischar~
continue until manually terminated per 0-01-77a, "Waste Collector/Surge flo System Processing;" Radwaste Discharge Isolation Valve does NOT auto close on high discharge flo Cl automatically terminate without any additional operator action; Radwaste Discharge Isolation Valve will auto close on LOW discharge flo (1 continue until manually terminated per 0-01-77a, "Waste Collector/Surge System Processing;" Radwaste Discharge Isolation Valve will ONLY auto close on high discharge radiatio Friday, February 29,20083:01 :05 AM 78
0610 NRC RO EXAM KIA Statement: 268000 Radwaste A2.01 - Ability to (a) predict the impacts of the following on the RADWASTE ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: System rupture KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the response of the RADWASTE system due to a rupture of a plant system and the procedures used to mitigate that conditio 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 outcom NRC Exam Friday, February 29,20083:01 :05 AM 79
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: Recognize that the given location of the rupture places it downstream of the Radwaste Discharge Isolation Valve. Using that knowledge, recognize the conditions which cause an automatic isolation of that valve. The Radwaste Discharge Isolation Valve (FCV-77-61) auto closes on radiation monitor upscale, downscale or INOP. The valve will ALSO close on low CCW flow or the specific unit's cooling tower gate "A" positio The same is true of the High or Low flowrate isolation valves downstream of the 77-61 valve. These valves are named "High" or "Low" based on the anticipated flowrate of the radwaste discharge, which is based on the chemistry sample taken prior to the discharge. Neither valve automatically closes on high or low flow. No provision is made to automatically isolate the radwaste discharge line based on a system rupture. This must be accomplished manuall correct: A - incorrect: The Radwaste Discharge Isolation Valve will NOT automatically close under the given condition C - incorrect: The Radwaste Discharge Isolation Valve will NOT automatically close ( under the given condition D - incorrect: Although the discharge must be manually terminated, the Radwaste Discharge Isolation Valve will automatically close on other conditions besides high radiatio Friday, February 29,20083:01 :05 AM 80
OPL 171.084 Revision 5 Page 41 of 63 Instructor Notes (2) Discharge Control Station (a) Purpose - used to control rate of release, direct to discharge canal or cooling tower blowdown, and provides system isolatio (b) Interlocks TP-6 FCV 77-61 - Caution Order is Radwaste discharge placed on isolation auto closure Assoc. Unit 77-on Radiation monitor 61 valve when in 2': upscale isolation helper mod setpoint, downscale, or Inop, Unit specific Obj. V. A gate is not full open, or two CCWP's are not operating i FCV 77-588 radwaste high flow rate discharge isolation valve auto closure on Radiation monitor 2': upscale isolation setpoint, downscale, or Ino All 3 unit 1A gates are closed and cooling tower blowdown flow is ~ 60,000 gp OPL 171.084 Revision 5 Page 42 of 63 Instructor Notes ii FCV 77 -58A radwaste low flow rate discharge isolation valve auto closure on (1 )Radiation monitor 2: upscale isolation setpoint, downscale, or Inop (2)AII 3 unit 1A gates closed, and Cooling tower blowdown flow < 60,000 gpm i FCV 77-279 radwaste isolation valve to cooling tower blowdown auto closure on (1 )Radiation monitor 2: upscale isolation setpoint, downscale, or Inop (2)cooling tower blowdown flow
~ 60,000 gpm (c) Flow rate is set by 0-81-4.8.A.1-1 and is controlled by FCV-77-59A or B by use of controls on panel 25-17.
(d) Recorders Flow recorder provides indication of release rate and permanent record of flow rate and total discharge release tim NRC RO EXAM 37. RO 272000K5.01 OOllCIA/SYS/HWCIB9/272000K5.01llRO/SRO/BANK Given the following plant conditions:
* The Hydrogen Water Chemistry (HWC) System is in the "Operator Determined Setpoint" mod * Hydrogen flow is set at 14 SCF Which ONE of the following describes the plant response IF reactor power is reduced?
Main Steam Line radiation levels will _ _---->C....,:.1../....)_ _ _ _ _ the lowering of reactor powerdueto _ _ _ _ _ _ _ _~(2~)___________ 1) A'! rise in opposition to; a rise in volatile Ammonia productio B. lower in response to; a reduction in Nitrogen concentratio C. lower in response to; a reduction in Hydrogen concentratio D. rise in opposition to; a rise in Nitrite and Nitrate productio KIA Statement: 272000 Radiation Monitoring K5.01 - Knowledge of the operational implications of the following concepts as they apply to RADIATION MONITORING SYSTEM: Hydrogen injection operation's effect on process radiation indications: Plant-Specific KIA Justification: This question satisfies the KIA statement by requiring the candidate to use plant conditions to determine the effect on radiation levels due to specific operating conditions of the Hydrogen Injection syste 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 outcom NRC Exam Friday, February 29,20083:01 :05 AM 81
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. The mechanism by which Hydrogen injection causes Main Steam Line (MSL) radiation levels to rise higher than norma . The effect on hydrogen concentration in the reactor at reduced feedwater flow with Hydrogen Injection flowrate unchange . The effect on ammonia production due to a reduction of oxygen concentration in the reacto A - correct: B - incorrect: This is plausible because nitrogen concentration DOES decrease with power level. However, due to the reduction in oxygen concentration, the reduction in nitrite and nitrate production allows more nitrogen available to combine with the excess hydrogen and form volitile ammoni C - incorrect: This is plausible because of the TYPICAL response of the HWC system when operated in Automatic mode. In addition, understanding that hydrogen injection flowrate is constant may not lead to an understanding of that relationship to a reduction in feedwater flow where hydrogen is injected.. However, in Operator Demand mode the hydrogen concentration increases due to the reduction in feedwater flo D - incorrect: This is plausible because MSL radiation levels DO rise in opposition to lowering power. However, the reduction in available oxygen causes a reduction in nitrite and nitrate production, which allows more volatile ammonia formation.
Friday, February 29,2008 3:01 :05 AM 82
OPL 171.220 Revision 4 Page 14 of 71 h. O2 source LP turbine (1 ) Air in-leakage - oxygen in the air blading to leaks into the low pressure parts discharge of of the steam cycle condensate pumps is less (2) Some air in-leakage is removed than atmospheric by the SJAE's but some pressure dissolves in the condensate (3) Radiolysis reaction- One way O2 is produced 2H 2 O nandyradiation . ) 2H + 0 2 2 O 2 removal (1 ) Radiation induced recombination Reaction is an of H2 and O 2 equilibrium reaction 2H 0 f-~nd gaml11a~iation ~ 2Hz.+ O (2) Carry-over with the steam Hydrogen addition (1 ) When an excess of hydrogen is This is the basis injected to the feed water, the of H2 injection reaction is drivento the left and less oxygen (and peroxide) is produced (2) The chemical environment becomes less oxidizing (3) Elements exposed to the coolant will assume chemical forms using less oxygen and/or more hydrogen (4) Solubility and volatility may be This is how MSL affected by the change in radiation levels
"oxidation state" of the element will increase which will be discussed later
OPL 171.220 Revision 4 Page 27 of 71 G. Operation Be very careful when selecting Normally controlled from the HWC Main functions from Control Panel different screen Use self When using the Operator Interface Unit checkin (OIU) function buttons, be aware that the same function key will cause different actions on different screens INPO SER 3-05 Operation of the HWC PLC (1 ) Hydrogen controller Flow controller operates in 2 (a) A(i(;Jif;malHi~k~biwe,f;\\ modes ait~'~,niij~~d*. ,~e:tpojrit*.MOGl.e,
- changes hydrogen Procedure Use injection flow in response to changes in reactor power. Used for normal ( operation of the HWC System and when reducing hydrogen injection related dose rates to support maintenance, chemistry or radcon activities while the plant is operating (b) .~fl~8J; Hydrogen flow D.~~~;rPl]i;Q~d. .* *.~E?~p{)iht;Mbd.e . .*. . stays constant,
- changes hydrogen ' regardless of injection flow in response power changes, to the setpoint being until operator manually entered by the manually enters a operator. Normally used new setpoint when initially pressurizing, purging and placing the HWC System in service or if Power Determined setpoint is unavailable ( (2) Oxygen controller - Only mode used Automatic/Hydrogen Determined for oxygen control Setpoint
OPL 171.220 Revision 4 Page 45 of 71 Supply Facility Trip - A shutdown signal is generated when either the hydrogen or oxygen gas supply facility trips Hydrogen or Feedwater Flow Signal Failed - A shutdown signal is generated when the hydrogen flow signal or the feedwater flow signal is less than 2 mA or greater than 22 mA I. Radiological Effects of HWC On MSL's Obj. V. Obj. V.BA The primary source of background MSL Obj. V. radiation levels during reactor operation is due Obj. V. to the decay of nitrogen-16 (N 16 ) N 16 has a half-life of 7.1 seconds A 6.13 or 7.12 Mev gamma is emitted 6.13 Mev gamma on N 16 decay is more common M~or source of nitrogen in a BWR is 0 16 (11,P) 0 16 + 11 -7 N16 + P N1 reaction When using normal water chemistry methods, a major portion of the N16 present in the reactor coolant combines with the free oxygen to form water soluble nitrites (N0 2) and nitrates (N0 3) These compounds are circulated through the reactor coolant systems and are ultimately removed by the RWCU System A smaller fraction of the N 16 is carried Predominate over in the steam in the form of nitrogen contributor to gas (N 2) and ammonia (NH3) background radiation levels H2 injection alters the N 16 carryover ratio Concentrations of N0 3, N0 2, and NO decrease Concentration of NH3 increases Ammonia*
OPL 171.220 Revision 4 Page 46 of 71 (1) A gas (2) High water solubility 5.
6. The increased dose rates are due to the increased ease with which N 16 gets out of the We can maintain reactor and into the steam pipes when in the up to 2.7 ppm NH3 form injection concentration.
7. The initial U2 run was the first week in No . Up to 90 scfm hydrogen was injecte MSL '8' was Average MSL radiation level increased highest at approximately 5 times normal times normal 8. Addition of noble metals to reactor water Noble metals decompose during reactor Rubidium and startup or shutdown Iridium During this time it produces a thin layer of noble metal on wetted surfaces The ECP on these surfaces are reduced significantly during subsequent operation This leaves a stoichiometric excess of hydrogen Now the amount of hydrogen injection can be reduced which will lower MSL radiation levels This NOTE
OPL 171.220 Revision 4 Page 53 of 71 Consequences of Event No effects were noted. However there is the potential for rapid recombination in the Offgas Charcoal beds. Additionally excessive hydrogen increases the risk for explosion or fire .. 2. High radiation on reduction in power event at Monticello Event description On December 13, 1997 with Monticello at approximately 75 percent power, workers entered the main condenser room to repair a leaking root valve and found the dose rate 2.5 times greater than expected Reactor power had been reduced from 100 percent power to 75 percent power for ALARA purposed and hydrogen water chemistry injection rate had been reduced from the normal 40 scfm to 8 scfm Dose rate encountered was significantly Encountered - higher that expected 4,800 mrem/hr Expected - 2,000 Job was stopped to evaluate the mrem/hr situation and management decided to have power reduced further At 60 percent power, dose rates were about 3,200 mrem/hr and the job was completed
OPL 171.220 Revision 4 Page 54 of71 b. Cause of event Lack of understanding of the Questioning radiological effect of reducing reactor Attitude could power under HWC conditions. As have prevented reactor power is decreased, less N-16 is thi produced, and steam line dose rates decrease. However, power level changes also change feedwater flow rate and hyd rogen concentration Reactor power and feedwater hydrogen concentration both affectsteam line dose rates. At a constant hydrogen injection rate, as power is decreased, feedwater flow rate decreases, and hydrogen concentration increases An increase in hydrogen concentration increases the ammonia concentration although hydrogen injection rates were reduced, the hydrogen concentration increase that occurred when feedwater flow rate decreased with reactor power was not accounted for and resulted in higher than expected dose rates c. Consequences of Event: Rx power needed to be lowered to 60% Work Planning vice 75% planned. This resulted in unplanned lost generatio Work had to be performed in a 3200mr/hr vice a planned 2000 mr/hr field. This resulted in unplanned exposur Potential lost opportunity to plan and perform work that required Rx power to be lowered to 60 % power vice the planned reduction to 75% power
0610 NRC RO EXAM 3 RO 290003A3.01 OOlIMEM/T2G2IHVAC/4/290003A3.01l3.3/3.5IRO/SROIBANK Given the following plant conditions:
* High radiation has been detected in the air inlet to the Unit 3 Control Roo * Radiation Monitor RE-90-259B is reading 250 cp Which ONE of the following describes the Control Room Emergency Ventilation (CREV) System response?
_ .......(1"'-')_ _ CREVS Unit(s) will automatically start _ _ _ _ _"""'(2=..)'--_ _ __
(1 ) (2) NEITHER; at the current radiation leve BOTH; with suction from the normal outside air path to Elevation 3 Selected; and will continue to run until Control Bay Ventilation is restarted; then, it will automatically sto D.oI Selected; Standby CREVS Unit will begin to auto-start; but, will ONLY run if the selected CREVS Unit fails to develop sufficient flo KIA Statement:
290003 Control Room HVAC A3.01 - Ability to monitor automatic operations of the CONTROL ROOM HVAC including: Initiationlreconfiguration KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the effect on CREV initiation logi References: Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam Friday, February 29, 2008 3:01 :05 AM 83
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: The Control Room Emergency Ventilation (CREV) system automatically starts on high radiation sensed by Radiation Monitor RE-90-259B. The actual stepoint is 221 cp The radiaf . . en in the stem was chosen because the Tech Spec In I his requires the candidate to determine the difference for operational validity. THere are two CREV units. One is selected as "Primary" and the other "Standby". On initiation, both CREV units receive a start signal, however the
"Standby" unit's signal is delayed 30 seconds to allow the "Primary" unit enough time to start. If the start is successful, the flow from the "Primary" unit will remove a start permissive to the "Standby" unit and it will abort it's startup. CREV units must be manually secured once the initiating conditions have cleare D - correct:
A - incorrect: This is plausible because the Tech Spec initiation setpoint is 270 cpm, which is less than the given radiation level. However, the actual CREV initiation setpoint is 221 cp B - incorrect: This is plausible since both CREV units receive a start signal on a valid initiation. However, the CREV unit NOT selected will experience a 30 second time delay on initiation and will only complete its start sequence if the selected CREV unit fails to star C - incorrect: This is plausible because the start sequence is correct. However, once initiated, CREV must be manually secured. There is no automatic shutdown capability, only trips.
Friday, February 29, 2008 3:01 :05 AM 84
OPL 171.067 Revision 13 Page 30 of 71 INSTRUCTOR NOTES 7. Control Room Emergency Ventilation (CREV) is Tech. Spec. 3. designed to supply and process the outdoor air ObjV.B.2/ V.B.5/ needed for pressurization during isolated V.C.6N. conditions. There are 2 CREV units rated at (Old CREV Units 3000 cfm each. A CREV unit consists of Motor- abandoned in place driven fan, (power supply is from 480V RMOV as Auxiliary Bd 1A for CREV Fan A; RMOV Bd 3B for CREV Pressurization Fan B), HEPA filter (common), charcoal filter Systems) assemblies located in the CREVS Equipment TP-4 Room, charcoal heater, and inlet isolation 2-47E2865-4 damper and a backflow check outlet dampe They are designed to maintain a positive pressurization to 1/8" w.g. minimum to the control roo A CREV may be started manually from Red indicating lights control room Panel 2-9-22 if local control on panel 3-9-21 to switch is in AUTO position via a 3 provide indication of position, spring-return to center switc CREV Fan A and/or (STOP-AUTO-START). Actuates only the B running on Unit CREVS unit & associated damper, not the Annunciators are on isolation damper panel 9-6 for all unit There is also a 2 position maintained contact, one per train, AUTO-INITIATE/ TEST switch which is used to perform system level actions for that train (primarily testing). It provides the same response as auto star Local start at local control station in relay room is done using a 2 position maintained, one per train, AUTO-TEST switch. Isolation dampers do not operate automatically if started from local pane Automatic start signals are: Obj. V.B.1N. (1) High radiation of 221 cpm above Obj. V. background (270 cpm Tech Specs) Obj. V.C.17 in air inlet ducts to control room from (Radiation monitor RE 90- T. S. 3.3. A Units 1 & 2, Radiation monitor RE 90-259B Unit 3). Either monitor starts selected CREV uni (2) Reactor zone ventilation systems radiation high ~72 MR/hr
OPL 171.067 Revision 13 Page 31 of 71 INSTRUCTOR NOTES (3) Refuel zone ventilation systems The inlet damper is radiation high "?.72 MR/hr normally closed & fails closed. Damper (4) Low reactor water level at +2 inches opening takes -70 above instrument zero seconds. While in (5) High primary containment pressure the intermediate
"?.2.45 psig position both red &
e. On receipt of a start signal, normal outside green lights will be lit air paths (see below) to elevation 3C are on 2-9-2 isolated. The selected CREV unit starts The unit heater will once the inlet damper is full open. This energize 10 se supplies pressurizing air to the Unit 1, 2 after the damper is and 3 control rooms. One CREV unit can full open to allow the supply all three control rooms, so the STBY fan to come up to CREV unit will not normally start. Once speed. High Rad or started, the CREV unit will continue to run PCIS signal will until manually secured by first clearing the energize relays in high radiation signals and the PCIS signals Div I (CR1-A) and (otherwise equipment cycling will occur) Div II (CR1-B).
Contacts from the f. Control bay (EL 617) isolation is CR 1 relays are used accomplished by five pneumatic and motor- to energize operated low leakage dampers which solenoids to isolate isolate all normal air intakes and exhausts the M.C.R. normal forEL61 intake dampers (1) FCO-31-150B, fresh air make-up (150B,D,E,F, and G) duct to Units 1 and 2 Control Room and Relay Room AH (2) FCO 31-150G, 3C elevation relief vent isolation (3) FCO-31-150E, exhaust from Unit 1 toilet, locker, and other rooms at elevation 61 (4) FCO-31-150D, mounted in fresh air makeup duct to Unit 3 Control Room AH (5) FCO-31-150F, exhaust from unit 3 toilet, locker, and other rooms at elevation 617.
(
OPL 171.067 Revision 13 Page 32 of 71 INSTRUCTOR NOTES Manual initiation of the emergency mode Adherence to of operation can be performed from the procedures control room by operation of the AUTO- INPO SER 03-05 INITIATEITEST switch (putting switch in Obj.v.B.4N. INITIATEITEST position) at Pnl 2-9-2 This operation results in energizing the CR 1 relay for that train/division and isolation of the control room damper Only one solenoid must be energized to close these dampers. Therefore, either test switch will initiate a damper isolatio h. One switch alone in the INITIATEITEST Normally "A" is position will NOT result in full functionality selected unit via of the CREVS units. If that train is not the switch in CREV selected unit, then operation of that train room. If "A" is will be delayed by approx. 30 seconds, inoperable, switch waiting for the selected train. This delay 0-XSW':031-7214 will result in the operator waiting to see SYSTEM PRIORITY the result of his operation of the switc SELECTOR The operator will see the amber light lit, SWITCH is placed in indicating energization of the CR 1 relay TRAI N B position to and the solenoid for isolation damper start it without time closing, but will see no activity of the dela CREVS unit until the delay timer has timed out. If the selected train's switch is put in the INITIATEITEST position that train will immediately enter its initiation sequence, with the damper's red light being lit as well as the green, indicating travel of the damper toward the open positio However, should there be any failure of the selected unit; the standby unit will not start. This is because the CR1 relay for the standby unit was not actuated. Therefore, when manually initiating emergency operation of the new CREVS units, it is important to put the AUTO-INITIATEITEST switches of BOTH trains to the INITIATEITEST positio OPL 171.067 Revision 13 Page 33 of 71 INSTRUCTOR NOTES k. Again, to secure operation, the AUTO-INITIATEITEST switches must both be returned to the AUTO position and then the STOP-AUTO-START switches turned to the STOP position, to .reset the CR1 relays in both divisions.
I. Trips for the units, which are effective at all times, are the following:
(1) Fan overload (2) Unit low flow, less than approx. 2700 cfm -- trip is delayed for 10 seconds after fan star (3) High heater discharge temperature, approx. 220°F (4) Low heater delta temperature(between unit inlet and heater discharge}, indicating that the heater is not getting the relative humidity below 70 % -- trip is delayed for approx. 15 seconds after the heater is energized.
m. When any of these trip signals are ObjV.BA/ V. received, the following will occur:
(1) The heater will be immediately deenergize (2) The fan will continue to run and the damper will remain open for appro seconds, to dissipate the heat from the heater. (I n the case oHan overload, the fan will trip immediately.)
(3) The inlet damper will be deenergized, and when no longer fully open, the fan will be deenergized. The damper requires approx. 20 seconds to close, while fan coast down is approx. 60-90 second OPL 171.067 Revision 13 Page 34 of71 INSTRUCTOR NOTES n. In addition to the trips shown above, loss of power to the inlet damper will trip the unit. In this case, the heater is immediately tripped and the fan is deenergized when the damper is no longer fully open. This action results in (slightly) faster tripping of the heater to avoid heat dissipation problems.
o. Flow switches are provided, one for each PDIS 7316 at Unit 2 division/unit, to start the standby unit if Vent Tower Intake the selected unit does not start or trips Plenum off. The selected unit not starting is sensed by low differential pressure across the common HEPA filter in the Unit 2 vent tower. Low differential pressure exists when a fan is not operating; this signal will normally be present. The circuit for each unit is such that its initiation sequence is begun upon either of the following:
(1) Unit is selected as primary unit and CR1 relay for that division is energize (2) Other unit is selected as primary unit, low differential pressure exists across the common HEPAfilter, and CR1 relay for that division has been energized for approx. 30 seconds.
p. With this circuit design, when an accident signal is initially received, the selected unit will enter its initiation sequence immediately and the other unit will enter its initiation sequence approx. 30 seconds later. Once the selected unit fan has been started (taking approx. 75 seconds-
- 70 for the damper and 5 for the fan). the low differential pressure signal will no longer be present in the standby unit circuitry and its damper will return to the fail-close positio OPL 171.067 Revision 13 Page 35 of 71 INSTRUCTOR NOTES q. If the selected unit fails to start properly, it will itself be turned off by the trips noted above, and the standby unit will continue in its initiation sequence. The time delay for startup of the standby unit will be selected to ensure that regardless of the primary unit failure, both fans will not be running at the same time.
r. If the selected unit starts properly, but then trips at a later time, the standby unit will only be missing the low differential pressure signal to receive its start signa The standby unit will start when the selected system has completed its shutdown process and the fan has been deenergized.
s. To secure from emergency operation, the high rad signals and the PCIS signals must first be cleared (otherwise equipment cycling will occur). These signals must be cleared on both divisions to not have the standby unit start up when the selected unit is secure. The STOP-AUTO-START switches in the control room should then be moved to the STOP position for both units. This will reset I deenergize the CR1 relays in both divisions, reopen the control room isolation dampers and remove the start signal from the operating CREVS uni The CREVS unit heater will then be deenergized, with the fan continuing to Obj. V. run and the damper held open for appro seconds, and the damper closing and the fan turned off as discussed earlie NRC RO EXAM 3 RO 295001G2.1.14 00l/MEM/TlGl/68 - RECIRC/21295001G2.1.141IRO/SROINEW 10116107 Given the following plant conditions: (
* You are the At-The-Controls (ATC) operator on Unit 1 * * Unit 1 is operating at full power when 1A Recirculation pump trippe * The Unit Supervisor has directed you to carry out the actions of 1-AOI-68-1A, "Recirc Pump Trip/Core Flow Decrease OPRMs Operable."
Which ONE of the following describes the required operator action(s) that CANNOT be carried out from your watch station? A'! Perform 1-SR-3.4.1 (SLO), "Reactor Recirculation System Single Loop Operation."
B. REFER TO ICS screens VFDPMPA(VFDPMPB) and VFDAAL(VFDBAL) to assist in determining the cause of the Recirc Pump trip/core flow decreas C. IMMEDIATELY take actions to insert control rods to less than 95.2% loadline AND REFER TO O-TI-464, "Reactivity Control Plan Development and Implementation."
D. CHECK parameters associated with the Recirc Drive and Recirc Pump/Motor 1A(1 B) on ICS and RECIRC PMP MTR 1A & 1B WINDING & BRG TEMPS, 1-TR-68-71 to determine the cause of the Recirc Pump tri KIA Statement: 295001 Partial or Complete Loss of Forced Core Flow Circulation / 1 & 4 2.1.14 - Conduct of Operations Knowledge of system status criteria which require the notification of plant personnel KIA Justification: This question satisfies the K/A statement by requiring the candidate to use specific plant conditions to determine the required actions which require notification of plant personnel outside of the control room due to a partial loss of Recirculation flo References: 1-AOI-68-1 Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam Friday, February 29,20083:01 :06 AM 85
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Which actions are required to be performed by 1-AOI-68- . Which of the actions determined from Item 1 above CANNOT be carried out by the ATC operato A - correct: This duty is carried out by Reactor Engineering once notified by the Unit Operato B - incorrect: This is plausible because the BOP operator or STA typically carry out this action. However, utilizing ICS screens is available at the ATC watch station and within his required dutie C - incorrect: This is plausible because the BOP operator typically inserts control rods while the ATC operator executes 1-AOI-68-1 and acts as Peer Checker if possible. However, manipulating control rods is part of the ATCwatch station dutie D - incorrect: This is plausible because the BOP operator or STA typically carry out this action. However, utilizing ICS screens is available at the ATC watch station and within his required duties.
Friday, February 29, 2008 3:01 :06 AM 86
BFN Recirc Pump Trip/Core Flow Decrease 1-AOI-68-1 A Unit 1 OPRMs Operable Rev.OOO2 Page 7 of 12 Subsequent Actions (continued) NOTE 1) Step 4.2[3] through Step 4.2[lS.3] apply to any core flow lowering even ) Power To Flow Map is maintained in 0-TI-24S, Station Reactor Engineer and on IC [3] IF Region I or" of the Power to Flow Map is entered, THEN (Otherwise N/A) IMMEDIATELY take actions to insert control rods to less than 95.2% load line AND REFER TO 0-TI-464, Reactivity Control Plan Development and Implementatio [4] RAISE core flow to greater than 45% in accordance with 1-01-6 o
[5] INSERT control rods to exit regions if NOT already exited AND REFER TO 0-TI-464, Reactivity Control Plan Development and Implementatio o NOTE The remaining subsequent action steps apply to a single Reactor Recirc Pump tri [6] CLOSE tripped Recirc Pump discharge valv o [7] MAINTAIN operating Recirc pump flow less than 46,600 gpm in accordance with 1-01-6 o [S] [NER/C] WHEN plant conditions allow, THEN, (Otherwise N/A)
MAINTAIN operating jet pump loop flow greater than 41 x 106 Ibm/hr (1-FI-6S-46 or 1-FI-6S-4S). [GE SIL 517] o
BFN Recirc Pump Trip/Core Flow Decrease 1-AOI-68-1 A Unit 1 OPRMs Operable Rev. 0002 ( Palle 8 of 12 Subsequent Actions (continued) CAUTION The temperature of the coolant between the dome and the idle Recirc loop should be maintained within 75°F of each other. If this limit cannot be maintained, a plant cool down should be initiated. Failure to maintain this limit and NOT cool down could result in hangers and/or shock suppressers exceeding their maximum travel range. [GE SIL 251, 430 and 517J
[9] IF Recirc Pump was tripped due to dual seal failure, THEN (Otherwise N/A) [9.1] VERIFY TRIPPED, RECIRC DRIVE 1A(1 B) NORMAL FEEDER, 1-HS-57-17(14). o [9.2] VERIFY TRIPPED, RECIRC DRIVE 1A(1 B)
ALTERNATE FEEDER, 1-HS-57-15(12). o
[9.3] CLOSE tripped recirc pump suction valve using, RECIRC PUMP 1A(1 B) SUCTION VALVE, 1-HS-68-1 (77). o [9.4] IF it is evident that 75°F between the dome AND the idle Recirc loop cannot be maintained, THEN COMMENCE plant shut down and cool down in accordance with 1-GOI-1 00-12 o [10] NOTIFY Reactor Engineer to perform Reactor Recirculation System Single Loop Operation, 1-SR-3.4.1 (SLO) AND to refer to Station Reactor Engineer, 0-TI-248 and Tech Specs 3.4.1 as necessar o [11] [NER/C] WHEN the Recirc Pump discharge valve has been closed for at least five minutes (to prevent reverse rotation of the pump) [GE SIL-517J, THEN (N/A if Recirc Pump was isolated in Step 4.2[9])
OPEN Recirc Pump discharge valve as necessary to maintain Recirc Loop in thermal equilibriu BFN Recirc Pump Trip/Core Flow Decrease 1-AOI-68-1A Unit 1 OPRMs Operable Rev. 0002 Page 9 of 12 Subsequent Actions (continued)
[12] REFER TO the following ICS screens to help determine the cause of recirc pump trip/core flow decreas * VFDPMPA(VFDPMPB) D * VFDAAL(VFDBAL) D [13] CHECK parameters associated with Recirc Drive and Recirc Pump/Motor 1A(1 B) on ICS and RECIRC PMP MTR 1A & 1 B WINDING & BRG TEMPS, 1-TR-68-71 to determine cause of tri D [14] PERFORM visual inspection of tripped Reactor Recirc Driv D [15] PERFORM visual inspection of Reactor Recirc Pump Drive relay boards for relay target D [16] IF necessary, THEN (Otherwise N/A)
REFER TO 1-01-68 for Reactor Recirc Pump trip D
[17] INITIATE actions required to make the necessary repair D NOTE Restarting a Recirc Pump while in Region 1 is NOT allowed. Tech Spec 3.4.1.A requires that the Reactor Mode Switch be immediately placed in SHUTDOWN upon entry into Region 1 [18] PERFORM the following for Single Loop Operation: [18.1] REFER TO 1-01-68 for guidance on single loop operatio D [18.2] REFER TO Tech Specs 3. D [18.3] WHEN available, THEN RETURN tripped Recirc Pump to service in accordance with 1-01-6 NRC RO EXAM 40. RO 295001AK3.01 OOllMEM/TlGlIRECIRC!I295001AK3.011IRO/SRO/BANK Given the following plant conditions: * Unit 3 is operating at 100% powe * The following alarm is received: - RECIRC LOOP A OUT OF SERVICE (3-XA-9-4A, Window 26)
Which ONE of the following describes the Reactor Water Level response? A." Rises initially due to swell, then returns to norma Lowers initially due to shrink, then returns to norma Rises initially due to swell and remains high due to a lower power leve Lowers initially due to shrink and remains lower due to the loss of core void KIA Statement: 295001 Partial or Complete Loss of Forced Core Flow Circulation I 1 & 4 AK3.01 - Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Reactor water level response KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the effect on reactor water level due to a partial loss of Recirculation flo References: Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam Friday, February 29,20083:01:06 AM 87
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. The internal response of the Reactor Vessel due to a Recirc Pump tri . The RPV level instrument response to the conditions determined from Item 1 abov The RPV response occurs in three part First, the trip of the recirc pump causes a sudden reduction of coolant flow up through the fuel bundles while power level remains approximately 100%. This causes a sudden increase in core void fraction. The large voiding in the active fuel region coupled with the reduction in inventory being removed from the downcomer by the tripped recirc pump results in a rapid rise in RPV level outside the core shroud while water level inside the core shroud lowers. Since RPV level is measured outside the shroud, indication rise Next, the large void content in the active fuel region responds quickly to insert negative reactivity, causing a large reduction in reactor power and therefore, steam (void) production. This reduction in void fraction draws water from the downcomer region outside the shroud into the active fuel region inside the shroud. Even though reactor power will drop to approximately 65% with a 100% rod pattern, core void fraction at 65% is actually greater than at 100% due to the effect of recirc flow. Therefore, RPV level indication does not immediately return to its original valu Finally, the Feedwater Control System responds to the transient by reducing feedwater flow below steam flow to enable RPV level to slowly return to the original setpoint at a lower reactor powe A - correct: B - incorrect: This is plausible because level inside the core shroud initially lowers, then returns to normal. However, RPV level is NOT measured inside the core shrou C - incorrect: This is plausible because RPV level initially rises in response to the recirc pump trip. However, the lower power level is compensated by automatic adjustments of Feedwater Contro D - incorrect: This is plausible because level inside the core shroud initially lowers, then returns to normal. In addition, the reduction of core voids is temporary. Final void fraction is actually higher. However, RPV level is NOT measured inside the core shroud.
Friday, February 29, 2008 3:01 :06 AM 88
0610 NRC RO EXAM 41. RO 295003AA2.01 001IMEM/TlGlIO-GOI-100-41IRO 295003AA2.01l1RO/SROINEW 10/16/07 Given the following plant conditions:
* All three units were at 100% rated power when 500KV PCB 5234 (Trinity 1 feed to Bus 1 Section 1) tripped and failed to auto clos * The signal which caused the Power Circuit Breaker (PCB) trip CANNOT be rese * The Chattanooga Load Coordinator has issued a Switching Order directing BFN to open Motor Operated Disconnect (MOD) 5233 and 5235 to isolate 500KV PCB 5234 for troubleshootin Which ONE of the following describes your response to this Switching Order and the basis for that response?
ENSURE the PK block for PCB 5234 is_ _-.l...:(1.../...)_ _ _ _ to prevent (2)
(1) (2)
A. removed; to prevent electrical arcing across the MOD contacts while being opene B~ removed; to prevent actuating the breaker failure logic and tripping the remainder of the PCBs on Bus C. installed; to prevent actuating the breaker failure logic and tripping the remainder of the PCBs on Bus D. installed; to prevent electrical arcing across the MOD contacts while being opened.
Friday, February 29, 2008 3:01 :06 AM 89
0610 NRC RO EXAM KIA Statement: 295003 Partial or Complete Loss of AC / 6 AA2.01 - Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: Cause of partial or complete loss of power KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the potential cause of a partial or complete loss of AC powe References: 0-GOI-300-4 Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. The function of the PK block and its relationship to the breaker failure logi . Recognize the CAUTION in 0-GOI-300-4 related to PK block remova B - correct: A - incorrect: This is plausible because electrical arcing across the MOD contacts is actually what causes the trip signal to be generated if the PK block is installed. The electrical arcing will occur with or without the PK block installed. However, it will not trip the 500KV breakers on the bus when it happens without the PK block installe C - incorrect: This is plausible since the wording is ALMOST identical to the CAUTION, however the PK block must be remove D - incorrect: This is plausible because electrical arcing across the MOD contacts is actually what causes the trip signal to be generated if the PK block is installed. The electrical arcing will occur with or without the PK block installed. However, it will not trip the 500KV breakers on the bus when it happens without the PK block installed.
Friday, February 29, 2008 3:01 :06 AM 90
BFN Switchyard Manual 0-GOI-300-4 Unit 0 Rev. 0065 Page 65 of 85 Response to a Breaker Trip on 161 kV or 500kV Breaker CAUTION Breaker reclosure times on opposite ends of the transmission lines leaving BFN are 15 to 17 seconds after a trip. The breakers at BFN should reclose immediately thereafte [1] IF a line trips, THEN WAIT 30 seconds before resetting the disagreement to ensure adequate time for automatic reclosur NOTE kV breakers have high speed and standard speed reclosur . For PCBs equipped with digital relays, the PCB will lockout from AUTO closure if the affected line does not reclose from the other end within approximately seconds. Only the AUTO closure is prevented, the breaker can be manually closed with Dispatcher concurrenc CAUTION Induced currents in the current transformers of a 500KV PCB during cycling of the associated MOD's, in conjunction with an existing PCB trip signal, may actuate the breaker failure logic and trip all PCB's on the associated 500KV bus. Thus the MOD's associated with a tripped PCB should NOT be operated until the trip has been reset; or, if the trip cannot be reset, the breaker failure PK block has been removed for the associated tripped PCB during MOD operation. Contact Dispatcher for instruction or assistance to reset the tripped rela NRC RO EXAM 42. RO 295004AKl.03 OOl/MEM/Tl G 1/24VDCIVB91295004AKl.03IIRO/SROIBANK Given the following plant conditions:
* A reactor startup is in progress on Unit 3 and reactor power is on IRM Range * The operator observes the following indications: - SRM DOWNSCALE ( 3-XA-9-5A, Window 6) - IRM CH 'A', 'C', 'E', and 'G' HI-HI I INOP (3-XA-9-5A, Window 33)
Which ONE of the following power sources, IF lost, would cause these failures? A. 01 +1-24V DC Power Distribution Panel +1-48V DC Power Distribution Panel V AC RPS Power Supply Distribution Panel V AC Instrument and Control Power Distribution Panel KIA Statement: 295004 Partial or Total Loss of DC Pwr 16 AK1.03 - Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: Electrical bus divisional separation KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the effect on a division of IRM instruments due to a loss of DC powe References: Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam Friday, February 29, 2008 3:01 :06 AM 91
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Which of th listed power supplies input to the IRM syste . Which power supply, if lost, would provide only those indications liste A - correct: B - incorrect: This is plausible because 48V DC supplies power to annunciator panels in the control room including the two annunciators listed in the ste However, other annunciators would also be affected by the loss that are not included on the lis C - incorrect: This is plausible because RPS supplies trip units associated with IRM However, the given annunciators are not indicative of the loads supplied by RP D - incorrect: This is plausible because 120V AC I&C Buses supply power to IRM detectors and drives. A loss of that power supply would also affect the entire division. However, the indications given in the stem are not indicative of loads supplied by I&C buses.
Friday, February 29, 2008 3:01 :06 AM 92
BFN Panel 9-5 1-ARP-9-5A Unit 1 1-XA-55-5A Rev. 0011 Page 42 of 43 Sensor/Trip Point: IRM CHAN B.D. Relay K16 A. Hi-Hi HI-HI/INOP 1. 116.4 on 125 scal B. INOP 1. Hi voltage lo . Module unplugge (Page 1 of 1) 3. Function switch NOT in operat . Loss of +/- 24 VDC to monitor.
Sensor Control Room Panel 1-9-12.
Location: Probable Flux level at or above setpoint.
Cause: One or more inoperable conditions exis Testing in progres Malfunction of senso Control rod drop accident.
Automatic A. Half-scram if one sensor actuates (except with Rx Mode Switch in RUN).
Action: B. Reactor scram if one sensor per channel actuates (except with Rx Mode Switch in RUN).
Operator A. STOP any reactivity change o Action: B. VERIFY alarm by multiple indication o C. RANGE initiating channel or BYPASS initiating channe REFER TO 1-01-92 o D. With SRO permission, RESET Half Scram. REFER TO 1-01-99 E. IF alarm is from a control rod drop, THEN REFER TO 1-AOI-85- F. [NRC/C] IF one or more IRM recorder reading is downscale, THEN CHECK for loss of +/- 24 VDC powe 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 86-40-03] 0 H. NOTIFY Reactor Enginee I. REFER TO Tech Spec Table 3.3.1.1-1, TRM Tables 3.3.4-1 and 3.3.5- References: 1-45E620-6 1-730E237-6, -10 1-730E915-10 1-730E915RF-12 1-SIMI-92B
BFN Panel 9-5 1-ARP-9-5A Unit 1 1-XA-55-5A Rev. 0011 Page 10 of 43 SensorlTrip Point: SRM Relay K-19 Count rate 5 cp DOWNSCALE (Page 1 of 1) Sensor Panel 1-9-12, MCR.
Location: Probable A. An un-bypassed SRM channel having a count rate :s; 3 counts per second.
Cause: B. SI (or SR) in progres C. Malfunction of sensor.
Automatic Rod block below range 3 on IRM and Rx Mode Sw. NOT in Run.
Action: Operator A. VALIDATE SRM downscal D Action: B. IF alarm valid, THEN REFER TO 1-01-92 during startup (Mode 2) operation or O-GOI-1 00-3A, -3C during refuel (Mode 5) operatio D C. NOTIFY Unit Superviso D D. REFER TO Tech. Spec. Sect. 3.3.1.2, Table 3.3.1.2-1, TRM Tables 3.3.4-1 and 3.3.5~ D References: 1-45E620-6-1 1-730E237-8
BFN Loss of I&C Bus A 1-AOI-S7 -SA Unit 1 Rev. 0042 Page S of 44 SYMPTOMS (continued) Loss of Main Steam Relief Valve position indicatio Loss of power to RCIC and HPCI Turbine Vibration circuitry and position indication for testable check valves, (Panel 9-3). Loss of RHRSW and EECW Division I instrumentation, (Panel 9-3, 9-20). Loss of SBGT A flow and differential pressure indication, (Panel 9-25). Loss of SLC A and B amber ready lights and valve position indication, (Panel 9-5). Loss of LPRM meter lights and APRM alarm lights, (Panel 9-5). Loss of Condensate - Feedwater and Heater Drains instrumentation, (Panel 9-6). Loss of SRM/IRM detector drive power and position indication, (Panel 9-5). Loss of one-half the blue scram lights and accumulator low pressure-high level light indications (Panel 9-5, 25-04). Loss of Control Bay Emergency Ventilation System Division Loss of AC Supply to +/- 24V NEUTRON MONITORING BATT CHGR A1-1 NEG SIDE, 1-CHGD-283-0000A1-1 and +/- 24V NEUTRON MONITORING BATT CHGR A2-1 POS SIDE, 1-CHGD-283-0000A2-1. (The Neutron Monitoring Battery System is rated to carry loads for 3 hours. STACK GAS CH1 RAD MON RTMR, 0-RM-090-0147B will be lost after this time period) Loss of Main Steam Line B; 0 and Feedwater Line B flow indicators and inputs to 3 Element Control and Rod Worth Minimize "B" Fuel Pool Demin valves: FCV-078-0063, FPC FlO OUTBD ISOL VLV Closes, FCV-078-0068, RX WELL INFL INBD VLV Closes, FCV-078-0066, FPC FlO 1A BYP VLV Open These actions result from loss of power to A and C skimmer surge tank low-low level switche I&C BUS A VOLTAGE ABNORMAL (1-XA-55-8C, Window 21).
( Short Cycle valves 1-FCV-002-0029A and1- FCV-002-0029B fail open due to loss of power to 1-FC-2-2 NRC RO EXAM 4 RO 295005AA1.04 001lCIAlTlG1Ii1295005AA1.041IRO/SROINEW 11128107 RMS Given the following plant conditions:
* Unit 1 is at 100% rated power when the Desk Unit Operator notices that number 3 Main Turbine Stop Valve (MTSV) position indication is reading 0%. * Numbers 1, 2, and 4 MTSV position indications all read 100%. * Maintenance investigation determines that the cause of the MTSV position indication failure is due to a mechanical failure of the Linear Variable Differential Transducer (L VDT).
- The Unit 1 Main Turbine receives a trip signa . Which ONE of the following describes the effect on Main Turbine operation and any required action?
Maill iOr6'Tii'8operation is _ _ _ _-';--_ _ _ _ _ _ _ _ _ _ _ _ _ _ __ af~cb~G The RPS logic contact for the #3 MTSV will NOT function so a turbine trip may not initiate a scra B. f'\JOT aFrected. The RPS logic contact is already OPEN for the #3 MTSV so a turbine trip will still initiate a scra C~ affQs:ffid. The Generator output breaker will NOT open on a turbine trip due to a 4-out-of-4 logic arragemen ~OT affested. The Generator output breaker will still open on a turbine trip due to a 2-out-of-4 logic arragement.
Friday, February 29, 2008 3:01 :06 AM 93
0610 NRC RO EXAM KIA Statement: 295005 Main Turbine Generator Trip I 3 AA 1.04 - Ability to operate andlor monitor the following as they apply to MAIN TURBINE GENERATOR TRIP: Main generator controls KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the required action following a Main Turbine Generator tri References: OPL 171.228 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 outcom NRC Exam REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Whether the LVOT position indicator feeds the RPS logic, the Turbine Trip logic, or bot . Based on the above answer, the affect of a turbine trip with the failure activ C - correct: A - incorrect: This is plausible because the Main Turbine operation is affecte However, the position indication supplied to RPS is a limit switch, NOT the LVOT position. Therefore, a turbine trip WILL initiate a scram signal to RP B - incorrect: This is plausible because the position of #3 MTSV is supplied to RPS logic. However, the position indication supplied to RPS is a limit switch, NOT the LVOT position. Therefore, RPS logic "sees" #3 MTSV as open until the turbine trip D - incorrect: This is plausible based on the different logic associated with the CIVs and RPS on the main turbine. However, the logic for MTSV inputs to open the generator output breaker is a 4-out-of-4 logic. Therefore, the generator output breaker will NOT automatically open.
Friday, February 29, 2008 3:01 :06 AM 94
OPL171.228 Revision 3 Page 63 of 81 INSTRUCTOR NOTES Consequences of Event This caused the Bypass valves to start opening. Due to the short duration of the error signal the bypass valves did not reach full open and subsequently close Operation of the bypass valves would impact Rx pressure, Rx power and Generator loa Corrective Action - EHC logic software was modified to eliminate the possibility of this. type response to a communications glitch.
2. At BFN on 1115/2006, the Unit 3 generator PER 95370 breaker failed to trip as expected on a turbine tri Description of Event The metal rod moves At BFN on 1115/2006, the Unit 3 to alter the magnetic generator breaker failed to trip as coupling of 2 expected on a turbine trip. The logic for opposing the generator breaker needs to see all the transformer stop valves closed and the CIV's closed secondary windings (either intercept or stop). The LVDT for to make an LVDT S.V#1 was failed such that the generator provide an output breaker would not open on a turbine tri proportional to the Operator action was taken to manually position of the metal trip the generator breaker ro OPL 171.228 Revision 3 Page 64 of 81 INSTRUCTOR NOTES Cause of Event Work practices The LVDT transformer coupling rod Monitor all became disconnected from the valve and parameters during a fell to a position which gave indication of transient and ensure
- 50% valve position. The affect on the automatic actions logic for tripping the generator PCB on a have occurred turbine trip was not recognized. Consequences of Event Tripping of the generator breaker on a turbine trip prevents a reverse power situation where the generator and turbine could attempt to rotate backwards, causing equipment damage. The unit operator's quick recognition and response to the breaker failure to trip prevented damag NRC RO EXAM 44. RO 29S006AK3.0S OOl/C/A/TlGI/RPS/1/29S006AK3.0SIIRO/SROIBANK Given the following plant conditions: * Power ascension is in progress on Unit 3 with the Main Turbine on lin * Control Rods are being withdrawn to increase powe * As reactor power approaches 35%, the Shift Technical Advisor (STA) notes that two (2) Turbine Bypass Valves are OPE Which ONE of the following describes the effect on the plant?
Regarding the UFSAR Chapter 14 analyses for a turbine trip, the above condition is (1) conservative than the assumptions used in the UFSAR because (2) more; it lowers the actual power level at which the RPS reactor scram on turbine trip is enable B." less; it raises the actual power level at which the RPS reactor scram on turbine trip is enable more; it lowers the peak vessel pressure for a design basis transient in regard to transition boilin less; it raises the actual power level for a design basis transient in regard to peak cladding temperature.
Friday, February 29, 2008 3:01 :06 AM 95
0610 NRC RO EXAM KIA Statement: 295006 SCRAM I 1 AK3.05 - Knowledge of the reasons for the following responses as they apply to SCRAM: Direct turbine generator trip: Plant-Specific KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the response to a Main Turbine trip and the basis for that response related to a reactor scra 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 outcom NRC Exam REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Which assumptions are of concern regarding the UFSAR analysis for a turbine tri . What affect the above conditions have on that analysi . Which thermal limit is of concern regarding the analyzed transien B - correct: A - incorrect: This is plausible because the initial power level prior to a scram is assumed in the analysis. However, the given conditions raise the initial power level which is less conservativ C - incorrect: This is plausible because RPV pressure affects transition boilin However, this is NOT the limit of concern during this analysis and the initial conditions are LESS conservative with regard to MCP D - incorrect: This is plausible because the condition is less conservative based on initil power. However, the limit of concern is NOT PCT, but MCPR.
Friday, February 29,20083:01 :06 AM 96
0610 NRC RO EXAM 45. RO 295016AA2.04 00l/CIA/TlGI/AOI-100-211295016AA2.04//RO/SROINEW 1211712007 RMS Given the following plant conditions: ( d due to a fir * Actions re being carried 0 in accordance with 3-AOI-1 00-2, Control Room Abandon en * RPV level at (+)20 inches and stead Reactor pressure is
* RHR Lo P I is in Suppression Pool Coolin In accordance with 3-AOI-100-2, "Control Room Abandonment", a Suppression Pool Temperature limit of less than or equal to J1.L°F has been established. The basis for this limit is _ _ _ _ _ _ _ _......(=2)1--_ _ _ _ _ _ _ _ _ _ _ _ _? (1 A. 95 of; to prevent exceeding the Technical Specification LCO before reaching Mode 4 (Cold Shutdown).
B. 110°F; to prevent exceeding the Heat Capacity Temperature Limit before the reactor can be verified to be shutdow C. 120 of; to prevent damage to the RCIC turbine from overheated lube oil, which is cooled by the Suppression Pool wate D~ 120 of; to prevent exceeding the design basis maximum allowable values for primary containment temperature or pressur Friday, February 29, 2008 3:01 :06 AM 97
0610 NRC RO EXAM KIA Statement: { 295016 Control Room Abandonment AA2.04 - Ability to determine andlor interpret the following as they apply to CONTROL ROOM ABANDONMENT: Suppression pool temperature KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the limitation and basis for Suppression Pool Temperature during a Control Room Abandonmen References: 3-AOI-100-2, Tech Spec Bases 3.6, EOIPM O-V-B 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 outcom NRC Exam REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. The Suppression Pool Temperature limit established by 3-AOI-1 00- . The basis for the above limi D - correct: A - incorrect: This is plausible because .:'5. 95°F is the normal operating limit imposed by Technical Specifications. However, this limit is NOT expected to be maintained during Control Room Abandonmen B - incorrect: This is plausible because the basis for.:'5. 110°F is correct. However, this limit is NOT expected to be maintained during Control Room Abandonment and the reactor is assumed to be shutdow C - incorrect: This is plausible because elevated SP temperatures can result in overheating the lube oil used for lubricating RCIC. This constitues the basis for Caution #6 in the EOls; but, does NOT apply during Control Room Abandonmen Friday, February 29,20083:01 :06 AM 98
BFN Control Room Abandonment 3-AOI-100-2 Unit 3 Rev. 0017 Page 16 of 90 Date _ __ Unit 3 Subsequent Actions (continued)
[15.7] ESTABLISH RHR system flow between 7,000 and 10,000 gpm as follows: o [15.7.1] MONITOR RHR SYS I TOTAL FLOW, 3-FI-74-79 at Panel 3-25-3 [15.7.2] THROTTLE OPEN 3-HS-074-0059C, RHR SYSTEM I TEST VL Vat 480V RMOV Bd 3A, Compt.12C, 0 [15.7.3] WHEN RHR SYS I TOTAL FLOW, 3-FI-74-79 indicates between 7,000 and 10,000 gpm, THEN DIRECT the operator to stop throttling 3-HS-074-0059 o [15.7.4] VERIFY CLOSED RHR SYSTEM I MINIMUM FLOW VALVE, 3-FCV-74-7, at either of the following: o * 480V RMOV Bd 3D, Compt. 4E, 3-BKR-074-0007 RHR SYSTEM I MINIMUM FLOWVLVFCV-74-7 (M010-16A), OR (Otherwise N/A) 0 * Rx Bldg - SW Quad - E1 541' local control switch RHR SYSTEM I MINIMUM FLOW VALVE, 3-HS-074-0007B. (Otherwise N/A) 0 [15.8] MONITOR SUPPR POOL TEMPERATURE, 3-TI-64-55B, at Panel 3-25-32 and MAINTAIN temperature less than 120°F , o
Suppression Pool Average Temperature B 3.6.2.1 BASES ACTIONS 0.1, 0.2, and 0.3 (continued) Additionally, when suppression pool temperature is > 110°F, increased monitoring of pool temperature is required to ensure that it remains:::; 120°F. The once per 30 minute Completion Time is adequate, based on operating experience. Given the high suppression pool average temperature in this Condition, the monitoring Frequency is increased to twice that of Condition A. Furthermore, the 30 minute Completion Time is considered adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal suppression pool average temperature conditio E.1 and If suppression pool average temperature cannot be maintained at:::; 120°F, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the reactor pressure must be reduced to < 200 psig within 12 hours, and the plant must be brought to at least MODE 4 within 36 hour The allowed Completion Time is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. ' Continued additio,! of heat to the suppression pool with suppression pool temperature> 120°F could result in exceeding the deSign basis maximum allowable values for primary containment temperature or pressure. Furthermore, if a blowdown were to occur when the temperature was> 120°F, the maximum allowable bulk and local temperatures could be exceeded very quickl (continued) BFN-UNIT 3 B 3.6-62 Revision 0
Suppression Pool Average Temperature B 3.6.2.1 BASES LCO b. Average temperature::;; 105°F when any OPERABLE IRM (continued) channel is > 70/125 divisions of full scale on Range 7 and testing that adds heat to the suppression pool is being performed. This required value ensures that the unit has testing flexibility, and was selected to provide margin below the 110°F limit at which reactor shutdown is required. When testing ends, temperature must be restored to ::;; 95°F within 24 hours according to Required Action A.2. Therefore, the time period that the temperature is > 95°F is short enough not to cause a significant increase in unit ris c. Average temperature::;; 110°F when all OPERABLE IRM channels are::;; 70/125 divisions of full scale on Range This requirement ensures that the unit will be shut down at> 110°F. The pool is designed to absorb decay heat and sensible heat but could be heated beyond design limits by the steam generated if the reactor is not shut dow Note that 70/125 divisions of full scale on I RM Range 7 is a convenient measure of when the reactor is producing power essentially equivalent to 1% RTP. At this power level, heat input is approximately equal to normal system heat losses.
APPLICABILITY In MODES 1, 2, and 3, a DBA could cause significant heatup of the suppression pool. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining suppression pool average temperature within limits is not required in MODE 4 or (continued) BFN-UNIT 3 B 3.6-59 Revision 0
OPERATOR CAUTIONS EOI PROGRAM MANUAL SECTION O-V-8
DISCUSSION: CAUTION #5 and CAUTION #6 CAUTION #5, this warns the operator of the potential plant response ifinjection of cold, unborated water into the core is too rapid under conditions where little or no margin to subcriticality may exist. This may result in a large increase in positive reactivity with a subsequent reactor power excursion large enough to substantially damage the core.
CAUTION #6, the HPCI and RCIC Lube Oil Coolers are cooled by routing part of the pwnp discharge fluid to the cooler. At elevated temperatures in the suppression pool, the turbine lube oil may get too hot to provide adequate lubrication. Only during EOT operations will the system be needed at such an extreme suppression pool temperature. Therefore, the EOls are an appropriate location for this caution .
- REVISION 2 PAGE 15 OF 15 SECTION O-V-B
0610 NRC RO EXAM 4 RO 295018AK2.01 001lMEM/TlGlIRBCCW/31295018AK2.01llRO/SROIBANK Which ONE of the following components would lose cooling upon isolation of the RBCCW NON-ESSENTIAL LOOP ISOLATION VALVE (2-FCV-70-48)? Recirculation Pump Seal Drywell Atmospheric Cooler C .... Spent Fuel Pool Cooling Heat Exchange Drywell Equipment Drain Sump Heat Exchange KIA Statement: 295018 Partial or Total Loss of CCW I 8 AK2.01 - Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER and the following: System loads KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific system knowledge to determine the effect on RBCCW loads due to a partial loss of RBCC References: 1/2/3-AOI-70-1, OPL 171.047 Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Which of the loads listed are part of the "Essential Loop".
2. Which of the loads listed are part of the "Non-essential Loop".
C - correct: A - incorrect: This is an "Essential Loop" loa B - incorrect: This is an "Essential Loop" load.
( D - incorrect: This is an "Essential Loop" loa Friday, February 29, 2008 3:01 :06 AM 99
BFN Loss of Reactor Building Closed 2-AOI-70-1 Unit 2 Cooling Water Rev. 0027 Page 14 of 14 Attachment 1 (Page 1 of 1) Components Cooled by RBCCW During Normal Plant Operation SYSTEM COMPONENTS COOLED Reactor Recirculation Pump Seals Pump Motor Bearings Pump Motor Windings Pump Discharge Sample Cooler Primary Containment Drywell Atmosphere Cooling Coils Reactor Water Cleanup Non-Regenerative Heat Exchangers Pump Seals Pump Bearings Fuel Pool Cooling and Cleanup Fuel Pool Heat Exchangers Equipment Drains Reactor Building Equipment Drain Sump Heat Exchanger Drywell Equipment Drain Sump Heat Exchanger
OPL 171.047 Revision 12 Page 10 of 41 Proper system flow operation is assured by Done Each Shift monitoring the system DP (pump discharge minus pump suction).
2. RBCCW Heat Loads Essential loop loads Obj. V. * Drywell Blowers(10) Obj. V. * Reactor recirculation pump motor coolers (2)
* Reactor recirculation pump seal coolers (2) * Drywell equipment drain sump heat exchanger (1) Non-essential loop loads Obj. V. * Reactor Building equipment drain Obj. V sump heat exchanger (1) * Reactor water cleanup pump seal water coolers and bearing oil coolers (2) * RWCU Non-regenerative heat exchangers (2) * Fuel pool cooling heat exchangers (2) * Reactor recirculation pump discharge sample cooler (1)
3. RBCCW Heat Exchangers These provide the means for heat removal DCN 51195, from RBCCW by RCW with Emergency replaced HX 1A & Equipment Cooling Water (EECW) as a 1B, HX 1C NOT backu replace OPL 171.051 They are counter-flow type, 50% capacity eac * RBCCW flow makes one pass through the shell sid * RCW makes one pass through the tube sid NRC RO EXAM 47. RO 295019AA2.02 001lCIA/SYS/CAI1295019AA2.021IRO/SROIBANK Given the following plant conditions:
* Unit 2 was at 100% power when a transient occurred which resulted in a reactor scra * After stabilizing the unit, the scram signal is RESE * All eight (8) Scram Solenoid Group lights are O * Approximately ten minutes later, the following conditions are present: - Raw Cooling Water (RCW) Low Pressure Alarm - CRD charging water High Pressure Alarm - Outboard MSIVs CLOSED, Inboard MSIVs are OPEN - Scram Discharge Volume (SDV) Vents and Drain Valves are CLOSED - Scram Solenoid Air Valves are OPEN Which ONE of the following describes the cause for the event?
A':" Loss of Control Ai B. Loss of both RPS busse C. Loss of Drywell Control Ai D. Loss of 9-9 cabinet 5, Unit Non-Preferred.
Friday, February 29, 2008 3:01 :06 AM 100
0610 NRC RO EXAM KIA Statement: 295019 Partial or Total Loss of Inst. Air / 8 AA2.02 - Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: Status of safety-related instrument air system loads (see AK2.1 - AK2.19) KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the effect of a loss of Control Air on safety related load References: 2-AOI-32-2 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 outcom NRC Exam REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Which indications given are indicative of the possible causes liste A* correct: B * incorrect: This is plausible because the scram would occur as well as scram valves open and SDV vents and drains closed, however these indications would NOT be appropriate AFTER the scram was reset. In fact, the scram could NOT be reset without RPS availabl C - incorrect: This is plausible because a loss of Drywell Control Air would cause MSIVs to close, however the INBOARD valves would clos D - incorrect: This is plausible because the ONLY indications given that would NOT apply would be outboard MSIVs closing and SDV vents and drains failure to re-open.
Friday, February 29, 2008 3:01 :06 AM 101
BFN Loss of Control Air 2-AOI-32-2 Unit 2 Rev. 0032 Page 5 of 25 PURPOSE This Abnormal Operating Instruction provides symptoms, automatic action, operator actions and expected system responses for loss of control air. SYMPTOMS AIR COMPRESSOR ABNORMAL annunciator, (1-XA-55-20B, Window 29) is in alar CONTROL AIR COMP G BKR ENERGIZED (0-XA-55-23B, Window 38) will reset (extinguish) when panel reset pushbutton is depresse CONTROL AIR COMP G MOTOR AMPS, 0-EI-32-2901, on Panel 1-9-20 indicates approximately zero amp Air Compressor G ICS Display shows Compressor G in an unloaded or shutdown conditio Air Compressor G ICS Display shows lowering Control Air Header Pressur Control Air Compressor G breaker trippe SERVICE AIR XTIE VLV OPEN (0-FCV-33-1 Open) annunciator, 0-PA-33-1A11(3) (Unit 1 and Unit 3) on Panel 1(3)-9-20 is in alarm at (1 (3)-XA-55-20B, Window 30). CONTROL AIR PRESS LOW annunciator, 0-PA-32-88 is in alarm (2-XA-55-20B, Window 32). SCRAM PILOT AIR HEADER PRESS lOW annunciator, 2-PA-85-38B on Panel 9-5 is in alarm (2-XA-55~5B, Window 28). MAIN STEAM LINE ISOL VLV POSN HALF SCRAM annunciator is in alarm (2-XA-55-4A, Window 30). DRYWELL CONTROL AIR PRESSURE LOW 2-PA-32-70 annunciator is in alarm (2-XA-55-3E, Window 35). CONDENSER A, B OR C VACUUM lOW 2-PA-47-125 annunciator is in alarm (2-XA-55-7B, Window 17). OG HOLDUP LINE INLET FLOW LOW 2-FA-66-111A annunciator is in alarm (2-XA-55-53, Window 4). HOTWELl A(B)(C) LEVEL ABNORMAL 2-LA-2-3(2-LA-2-6)(2-LA-2-9) is in alarm (2-XA-55-6A, Window 5(6)(7)).
BFN Loss of Control Air 2-AOI-32-2 Unit 2 Rev. 0032 Page 6 of 25 SYMPTOMS (continued) . REACTOR WATER LEVEL ABNORMAL 2-LA-3-53 ahnunciator is in alarm (2-XA-55-5A, Window 8). REACTOR PRESS HIGH 2-PA-3-53 annunciator is in alarm (2-XA-55-5A, Window 1). REACTOR CHANNEL A(B} AUTO SCRAM annunciator in alarm if any scram setpoint is exceeded (2-XA-55-5B, Window 1(2)). MAIN STEAM RELIEF VLV AIR ACCUM PRESS LOW 2-PA-32-31 annunciator in alarm (2-XA-55-3D, Window 18). AUTOMATIC ACTIONS Unit 2 to Unit 3 Control Air Crosstie, 2-PCV-032-3901, will close when Control Air Header pressure reaches 65 psig and lowering at the valv U-1 TO U-2 CO NT AIR CROSSTIE, 1-PCV-032-3901, will close to separate Units 1 and 2 when Control Air Header pressure reaches 65 psig and lowering at the valv PCV-84-0654, CAD/CA FLOW SEL, will select nitrogen from CAD tank A to supply 2-FSV-64-20, 2-FSV-64-21, 2-FSV-64-221, and 2-FSV-64-222 at
~75 psi PCV-84-0033, will select nitrogen from CAD tank A to supply 2-FSV-84-19, 2-FSV-64-29, and 2-FSV-64-3 PCV-84-0034, will select nitrogen from CAD tank B to supply 2-FSV-84-20, 2-FSV-64-31, and 2-FSV-64-3 BFN Loss of Control Air 2-AOI-32-2 Unit 2 Rev. 0032 Page 7 of 25 OPERATOR ACTIONS NOTE [NER/C] Attachment 1 provides expected system responses, critical components that do not fail in intended positions should be placed in the required positions. [INPO SOER 88-001] Immediate Actions None Subsequent Actions [1] IF a RFP Minimum Flow Valve failed open and flow is required from the condensate/feedwater system to reactor vessel or to prevent pump overload, THEN ISOLATE the associated RFP minimum flow lines in the appropriate RFPT Room as follows: (N/A any RFP valves not affected.)
- RFP 2A MIN FLOW SHUTOFF, 2-SHV-003-0508 o
* RFP 2B MIN FLOW SHUTOFF, 2-SHV-003-0517 o * RFP 2C MIN FLOW SHUTOFF, 2-SHV-003-0526 o [2] IF CNDS BSTR PUMPS DISCH BYPASS TO COND B, 2-FCV-2-29A and CNDS BSTR PUMPS DISCH BYPASS TO COND C, 2-FCV-2-29B fail CLOSED, THEN (Otherwise N/A) * VERIFY a flow path for condensate system OR * STOP the condensate pumps/booster pumps using 2-01- o [3] IF any outboard MSIVs fails closed, THEN:
PLACE associated hand-switch on Panel 2-9-3 to close position. (Otherwise N/A) o
BFN Loss of Control Air 2-AOI-32-2 Unit 2 Rev. 0032 Page 8 of 25 Subsequent Actions (continued)
[4] IF RSW STRG TNK ISOLATION VALVE, 0-FCV-02S-0032 FAILS CLOSED, THEN START a high pressure fire pump using 0-01-26 . o [S] OPEN CAD SYSTEM A N2 SHUTOFF VALVE, 0-FCV-084-000S, at PaneI9-S o [6] OPEN CAD SYSTEM B N2 SHUTOFF VALVE, 0-FCV-084-0016, at Panel 9-S o NOTES 1) All RCW temperature control valves fail open except for 2-TCV-24-80B and 2-TCV-24-8SB on 2A and 2B RBCCW heat exchangers and 2-TCV-024-007SB on the Main Turbine Oil Coolers (4" line) which fail close ) The appropriate computer pOints may be used for monitoring for the following lube oil temperatures, or any local temperature monitoring device that may be available, as necessary [7] IF RCW pump motor amps indicate that RCW System flow reduction is required, THEN REDUCE RCW flows as required: (Otherwise N/A). o [7.1] CLOSE main turbine lube oil cooler TCV isolation valve 2-SHV-024-0S83 or 2-SHV-024-0S84, THEN ESTABLISH lube oil temperature between 80°F and 90°F using TCV BYPASS VALVE 2-BYV-024-0S8S or 2-8YV-024-0S8 [7.2] CLOSE the following RFP turbine oil cooler TCV isolation valves * A RFP 2-24-624A or 2-24-62SA D* * B RFP 2-24-624B or 2-24-62SB o * C RFP 2-24-624C or 2-24-62SC o
BFN Loss of Orywell Control Air 2-AOI-32A-1 Unit 2 Rev. 0021 Page 4 of 9 PURPOSE This abnormal operating instruction provides symptoms, automatic actions and operator actions for the loss of Drywell Control Air System for causes other than Group 6 Isolation. The loss of Drywell Control Air caused by a Group 6 Isolation is addressed in 2-AOI-64-2d. SYMPTOMS DRYWELL CONTROL AIR PRESS LOW (2-XA-55-3E, Window 35) at
::;; 87 psi MAIN STEAM RELIEF VLV AIR ACCUM PRESS LOW (2-XA-55-3D, Window 18) at::;; 82 psi Inboard MSIV's close or start to close . . Drywell cooler dampers close. AUTOMATIC ACTIONS None
0610 NRC RO EXAM 48. RO 295021G2.4.50 OOllCIA/TlG1174-1112950212.4.501IRO/SRO/BANK Given the following plant conditions:
* Unit 2 is aligned with RHR Loop I in Shutdown Cooling and RHR Loop" in standby readines * A leak occurs in the RPV, which results in the following conditions: - RPV level at 0 inches and slowly lowering - Drywell Pressure at 3.0 psig and slowly rising - RHR Pumps 'A' and 'C' TRIPPED Which ONE of the following describes the minimum actions required to align RHR Loop " for injection to the RPV?
A. ... After FCV-74-47 OR FCV-74-48 is closed, push the RHR SYS " SD CLG INBD INJECT ISOL RESET 2-XS-74-13 After FCV-74-47 AND FCV-74-48 are closed, start RHR Loop" pumps, reset PCIS, and open the inboard injection valv After FCV-74-47 OR FCV-74-48 is closed; reset PCIS, push the RHR SYS " SD CLG INBD INJECT ISOL RESET 2-XS-74-132, and open the inboard injection valv After FCV-74-47 AND FCV-74-48 are closed, reset PCIS, push the RHR SYS " SD CLG INBD INJECT ISOL RESET 2-XS-74-132, and open BOTH injection valves.
Friday, February 29,20083:01 :06 AM 102
0610 NRC RO EXAM KIA Statement: 295021 Loss of Shutdown Cooling 14 2.4.50 - Emergency Procedures I Plan Ability to verify system alarm setpoints and operate controls identified in the alarm response manual KIA Justification: This question satisfies the KIA statement by requiring the candidate to analyze plant conditions and determine the required actions during an emergency which have resulted in a loss of shutdown coolin References: 2-AOI-74-1 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 outcom NRC Exam REFERENCE PROVIDED: None Plausibility Analysis: ( In order to answer this question correctly the candidate must determine the following: 1. Current RHR Loop II status given the initial condition . Based on the RHR Loop II status, determine the minimum actions to align Loop II for injection to the RP A - correct: B - incorrect: This is plausible because the valve alignment is correct and resetting 2-XS-74-132 is correct. However, resetting PCIS, re-opening FCV 74-47 and re-starting RHR pumps are NOT require C - incorrect: This is plausible because the valve alignment is correct and resetting 2-XS-74-132 is correct. However, resetting PCIS and re-opening FCV 74-47 is NOT require D - incorrect: This is plausible because the valve alignment is correct and resetting 2-XS-74-132 is correct. However, resetting PCIS and re-opening FCV 74-47 & 48 are NOT require Friday, February 29, 2008 3:01 :06 AM 103
BFN Loss of Shutdown Cooling 2-AOI-74-1 Unit 2 Rev. 0032 ( Page 7 of 31 Subsequent Actions (continued)
[5] IF Shutdown Cooling isolates on low RPV water level or high Drywell press (GROUP 2 ISOL) AND RPV water level needs restoring using LPCI, THEN (Otherwise N/A)
PERFORM the following before reaching -122 inches RPV water level: o NOTE The LPCI inboard injection valve that is aligned per 2-POI-74-2 will already be in the required accident position with the breakers open and will NOT isolat [5.1] PERFORM the following on a group 2 isolation: o
[5.1.1 ] IF 2-POI-74-2 is in effect, THEN VERIFY CLOSED one of the following valves: (Otherwise N/A) o * RHR SHUTDOWN COOLING SUCT OUTBD ISOL VLV, 2-FCV-74-4 * RHR SHUTDOWN COOLING SUCT INBD ISOL VLV, 2-FCV-74-4 o * VERIFY CLOSED the LPCI inboard injection valve NOT aligned for 2-POI-74-2, (RHR SYS I LPCIINBD INJECT VALVE, 2-FCV-74-53 OR RHR SYS II LPCI INBD INJECT VALVE, 2-FCV-74-67) 0
BFN Loss of Shutdown Cooling 2-AOI-74-1 Unit 2 Rev. 0032 ( Page 8 of 31 Subsequent Actions (continued)
[5.1.2] IF 2-POI-74-2 is NOT in effect, THEN VERIFY CLOSED the following valves on a Group 2 isolation: D * RHR SHUTDOWN COOLING SUCT OUTBD ISOL VLV, 2-FCV-74-4 D * RHR SHUTDOWN COOLING SUCT INBD ISOL VLV, 2-FCV-74-4 D * RHR SYS I LPCIINBD INJECT VALVE, 2-FCV-74-5 D * RHR SYS II LPCIINBD INJECT VALVE, 2-FCV-74-6 D [5.2] DEPRESS RHR SYS 1(11) SO CLG INBD INJECT ISOL RESET, 2-XS-74-126 and 2-XS-74-132 AND VERIFY 2-IL-74-126 and 2-IL-74':'132 extinguishe D
0610 NRC RO EXAM 49. RO 295023AK1.02 001lC/A/TlGlI79-21V.B.3.BI295023AK1.021IRO/SRO/MODIFIED 11117/07 Fuel loading is in progress on Unit 1 when you notice an unexplained rise in Source Range Monitor (SRM) count rate and an indicated positive reactor perio Which ONE of the following actions is an appropriate response? Immediately EVACUATE all personnel from the refuel floo If unexpected criticality is observed following control rod withdrawal, manually SCRAM the reacto C.'" If the reactor cannot be determined to be subcritical, traverse the refueling bridge and fuel assembly away from the reactor core, preferably to the area of the cattle chut If all rods are not insertedlcannot be inserted, verify the fuel grapple is latched onto the fuel assembly handle and immediately remove the fuel assembly from the reactor cor KIA Statement: 295023 Refueling Acc Cooling Mode I 8 ( AK1.02 - Knowledge of the operational implications of the following concepts as they apply to REFUELING ACCIDENTS: Shutdown margin KIA Justification: This question satisfies the KIA statement by requiring the candidate to analyze specific plant conditions to determine a reduction in Shutdown Margin has occurred and the actions required to address that conditio References: 1-AOI-79-2 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 outcom NRC Exam MODIFIED FROM OPL 171.060 #1 Friday, February 29,20083:01 :06 AM 104
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. The appropriate condition and Immediate Action required by 1-AOI-79- C - correct: A - incorrect: This is plausible because the evacuation of the Refuel Floor MAY be directed, but other actions to mitigate the problem take precedence until personnel safety is compromise incorrect: This is plausible because the condition is correct, but the action to scram is incorrect. Reinserting the control rod is require D - incorrect: This is plausible because the required action is correct, but the condition is NOT correct. This action is based on unexplained criticality following insertion of a fuel assembl ( ( Friday, February 29, 2008 3:01 :06 AM 105
BFN Inadvertent Criticality During Incore 1-AOI-79-2 Unit 1 Fuel Movements R~v.OOOO Page 6 of 9 OPERATOR ACTIONS Immediate Actions
[1] IF unexpected criticality is observed following control rod withdrawal, THEN REINSERT the control ro D [2] IF all control rods can NOT be fully inserted, THEN MANUALL Y SCRAM the Reacto D [3] IF unexpected criticality is observed following the insertion of a fuel assembly, THEN PERFORM the following: [3.1] VERIFY fuel grapple latched onto the fuel assembly handle AND IMMEDIATELY REMOVE the fuel assembly from the Reactor cor D [3.2] IF the Reactor can be determined to be subcritical AND no radiological hazard is apparent, THEN PLACE the fuel assembly in a spent fuel storage pool location with the least possible number of surrounding fuel assemblies and LEAVE the fuel grapple latched to the fuel assembly handl D [3.3] IF the Reactor can NOT be determined to be subcritical OR adverse radiological conditions exist, THEN TRAVERSE the Refueling Bridge and fuel assembly away from the Reactor core, preferably to the area of the cattle chute and CONTINUE at Step 4.1 [4]. D [4] IF the Reactor can NOT be determined to be subcritical OR adverse radiological conditions exist, THEN EVACUATE the refuel floo D
0610 NRC RO EXAM 50. RO 295024G2.1.33 OOIlC/A/TlGIICONTIPRIlBl0I295024G2.1.33//RO/SROIBANK During operation at 100% power a gross failure of both seals on Recirculation Pump 'B' increases Drywell Pressure to 2.0 psi Which ONE of the following is the approximate amount and type of Reactor Coolant System (RCS) leakage resulting from this condition? gpm of Identified leakage gpm of Unidentified leakage gpm of Identified leakage D." 60 gpm of Unidentified leakage KIA Statement: 295024 High Drywell Pressure I 5 2.1.33 - Conduct of Operations Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications KIA Justification: This question satisfies the KIA statement by requiring the candidate to determine that entry into Technical Specifications is required based on conditions which have resulted in high drywell pressur References: U2 TSR Sections 1 & 3.4.4, 2-AOI-68-1, OPL 171.007 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 outcom NRC Exam Friday, February 29, 2008 3:01 :06 AM 106
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Whether the leakage is IDENTIFIED or UNIDENTIFIED leakag . The amount of leakage associated with a gross failure of both seals on a single recirc pum D - correct: A - incorrect: This is plausible because the amount of leakage is correct. However, the leakage is not IDENTIFIED because the leakage is NOT intentionally captured and directed to a sump and is NOT expecte B - incorrect: This is plausible because the leakage is UNIDENTIFIED and equal to the Tech Spec value for total leakage, but insufficient for the conditions give C - incorrect: This is plausible because the leakage is equal to the Tech Spec value for total leakage, but insufficient for the conditions given. In addition, the leakage is not IDENTIFIED because the leakage is NOT intentionally captured and directed to a sump and is NOT expected.
Friday, February 29, 2008 3:01 :06 AM 107
RCS Operational LEAKAGE 3.4.4 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 RCS Operational LEAKAGE LCO 3. RCS operational LEAKAGE shall be limited to: a. No pressure boundary LEAKAGE; b. s 5 gpm unidentified LEAKAGE; and c. s 30 gpm total LEAKAGE averaged over the previous 24 hour period; and d. s 2 gpm increase in unidentified LEAKAGE within the previous 24 hour period in MODE 1.
APPLICABI LlTY: MODES 1, 2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Unidentified LEAKAGE Reduce LEAKAGE to 4 hours not within limi within limit OR Total LEAKAGE not within limi B. Unidentified LEAKAGE Reduce LEAKAGE 4 hours increase not within limi increase to within limit OR (continued) BFN-UNIT 2 3.4-9 Amendment No. 253
Definitions .1 Definitions (continued) LEAKAGE LEAKAGE shall be: a. Identified LEAKAGE 1. LEAKAGE into the drywell, such as that from pump seals or valve packing, that is captured and conducted to a sump or collecting tank; or 2. LEAKAGE into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; b. Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE; c. Total LEAKAGE Sum of the identified and unidentified LEAKAGE; d. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.
LINEAR HEAT The LHGR shall be the heat generation rate per unit GENERATION RATE length of fuel rod. It is the integral of the heat flux (LHGR) over the heat transfer area associated with the unit lengt (continued) BFN-UNIT 2 1.1-4 Amendment No. 253
BFN Panel 9-4 2-ARP-9-4A Unit 2 2-XA-55-4A Rev. 0032 Page 22 of 44 Sensor/Trip Point: RECIRC PUMP A 2-FIS-068-0055 0.1-0.2 gpm after second sea NO.2 SEAL LEAKAGE HIGH 2-FA-68-55 (Page 1 of 2) Sensor Recirculation Pump 2A Drywell Location: Probable A. Recirculation Pump 2A No.2 (outer) seal failure.
Cause: B. Sensor malfunction.
Automatic None Action: Operator A. COMPARE No.2 cavity pressure indicator (2-PI-68-63A) to No.1 Action: cavity pressure indicator (2-PI-68-64A), on Panel 2-9-4 or ICS. N seal degradation is indicated if the pressure at No.2 seal is less than 50% of the pressure at No.1 sea o B. IF seal failure is indicated, THEN INITIA TE seal replacement as soon as possible. Continued operation is permissible if Drywell leakage is within T.S. limit o NOTE 1) Possible indications of dual seal failure include:
* Window 25 on this panel alarming in conjunction with this windo * Rising drywell pressure and/or temperatur * Increased leakage into the drywell sum * Increased vibration of the recirc pum C. IF dual seal failure is indicated, THEN 0 1. SHUTDOWN Recirc Pump 2A by DEPRESSING RECIRC DRIVE 2A SHUTDOWN, 2-HS-96-1 . VERIFY TRIPPED, RECIRC DRIVE 2A NORMAL FEEDER, 2-HS-57-1 . VERIFY TRIPPED, RECIRC DRIVE 2A ALTERNATE FEEDER, 2-HS-57-1 . CLOSE RECIRC PUMP 2A SUCTION VALVE, 2-HS-68- Continued on Next Page
OPL 171.007 Revision 22 Page 17 of 86 ( (b) The flow keeps number 1 seal cavity clean and cool by flowing out of the seal area, along the pump shaft, and into the recirculation syste (c) This purge flow reduces the possibility of seal damage due to foreign material entering the seal from an unclean piping syste (8) Seal Failures Obj. V. Obj. V. (a) Seal failure may be assessed TP-6 by the resulting changes in flows and pressure (b) Failure of the number 1 seal ARPs provide useful assembly would allow a info/analysi higher flow to the number 2 Obj. V.D.2c seal cavity, forcing the Obj. V.E.3c number 2 seal to operate at a higher pressure (i.e., greater than 500 psig).
(c) This failure of the number 1 seal will cause leakage through the controlled seal leak-off line to rise to approximately 1.1 gpm. A flow element in this line causes a common alarm on high flow at 0.9 gpm or on low flow at 0.5 gp (d) Failure of the number 2 seal assembly would cause its seal pressure to drop (depending upon the magnitude of the failure).
(i) This failure would also cause a higher leakage through the seal leak detection line downstream from the number 2 sea OPL 171.007 Revision 22 Page 18 of 86 (ii) Normally there is no flow through this line and flow switches are set to alarm at 0.1- gpm flo (e) Failure of both mechanical Would cause seals would result in a total elevated drywell temp seal assembly leakage of 60 and pressure, and gpm as limited by the seal would exceed Tech breakdown bushing Spec and EPIP limits for RCS leakag (f) Should the number 1 seal restricting orifice become plugged, the RECIRC PUMP A(B) NO.1 SEAL LEAKAGE ABN annunciator will alarm on low flow (less than or equal to 0.5 gpm). Additionally, a reduction in number 2 seal pressure would be see (g) Should the number 2 restricting orifice become plugged, the RECIRC PUMP A(B) NO.1 SEAL LEAKAGE ABN annunciator would also alarm on low flow; however, number 2 seal pressure would rise to near the pressure of number 1 seal.
(9) Seal Cooling Obj. V. (a) Cooling for the recirculation Obj. V. pump seals is required due to Obj. V.B.19c the heat generated by the friction of the sealing surfaces and the leakage of reactor water through the seal assembl (b) This cooling is provided by a combination of supplied Reactor Building Closed Cooling Water (RBCCW) and the leakage of primary coolant past the seal .
0610 NRC RO EXAM 5 RO 295025EK2.08 OOl/C/A/TlG1IEHC LOGICI1295025EK2.08//RO/SROIBANK Unit 2 has experienced an inadvertent Main Steam Isolation Valve (MSIV) closure and subsequent reactor scram. Consequently, RCIC was placed in level control and is also maintaining reactor pressure 900 to 1000 psig with the MSIVs still isolate Given these plant conditions, the digital Electro-Hydraulic Control (EHC) System is in _--->.(..:...1)L-_ Pressure Control mode with the pressure setpoint set at _ ......(=2.1-)_ psi (1) (2) Reactor; 700 B." Header; 700 Reactor; 970 Header; 970 KIA Statement: 295025 High Reactor Pressure I 3 EK2.08 - Knowledge of the interrelations between HIGH REACTOR PRESSURE and the following: Reactorlturbine pressure regulating system: Plant-Specific KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the response of the digital EHC system to a transient resulting in a high reactor pressur References: OPL 171.228 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 outcom NRC Exam Friday, February 29, 2008 3:01 :06 AM 108
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Whether the header pressure dropped sufficiently low enough to cause an automatic transfer to Header Pressure Contro . Whether current plant conditions have allowed EHC logic to automatically transfer back to Reactor Pressure Contro B - correct: A - incorrect: This is plausible because EHC automatically transfers back to Header pressure control if HEADER pressure returns above 725 psig. However, the MSIVs are still closed so the given reactor pressure is not being sensed by the header pressure instrument C - incorrect: This is plausible because this condition is typical for post-scram EHC conditions if the MSIVs are ope D - incorrect: This is plausible because EHC will swap to Header pressure control, but the setpoint will drop from 970 psig to 700 psig.
Friday, February 29, 2008 3:01 :06 AM 109
OPL 171.228 Revision 3 Page 21 of 81 ( INSTRUCTOR NOTES 2. During turbine start-up and for a brief time Monitor Plant following synchronization, the bypass valve parameters for control also maintains the reactor steam expected response pressure. Once all the bypass valves are closed, then the turbine control maintains reactor steam pressure either in Header Pressure or Reactor Pressure Control depending on which operating mode is selecte . Steam pressure control is selectable from either Obj.V.B. panel 9-7 or the EHC Workstation by selecting HEADER PRESSURE CONTROL or REACTOR PRESSURE CONTRO . Header Pressure Control Input Signal Two redundant pressure transmitters Powered from within sense header pressure at the main steam the EHC system throttle just upstream of the main turbine stop valve Both signals are monitored for low, high, difference, and hardware failure The higher of the two Signals when no failures are detected is selected as the inpu A maximum difference setpoint of 10-PSI is also established to detect a fault and/or transmitter drift from either of the input In the event a fault is detected, the Obj.V.B. channel is prohibited from being used in the signal processing and the appropriate BYPASS pushbutton light will illuminate on 9-7 and on the HMI operator interfac Once the failed signal is corrected, depressing the BYPASS pushbutton will reset the BYPASS logic and both input signals will then be processe OPL 171.228 Revision 3 Page 22 of 81 INSTRUCTOR NOTES This mode IS NOT single failure proof - one of the two pressure sensors failing upscale can, and generally will be selected by the logic to control. This will open the TCV's and 8PV's to depressurize the header to the MSIV isolation setpoint of 852 psig in RUN Mod In the unlikely event that both inputs signals are detected as failed, the control logic will automatically switch to reactor pressure contro If header pressure drops below 700-PSI, and reactor pressure control is the controlling mode of operation, t' t 01 I I
. *If desired, the operator may re-select reactor pressure control after the transfer has been made even though header pressure is below 700-ps The automatic transfer logic will re-engage if header pressure rises above 725-psi.
5. Reactor Pressure Control Input Signal TP-3 Four (4) redundant pressure transmitters (PT- 204a-d) grouped in pairs with "A" and "8" constituting one pair and "C" and
"0" the other pai A pressure-biasing algorithm determines the lagged high-median value of the four (4) inputs and biases the remaining three (3) input signals to that high median valu The high-median Signal is then averaged Four biased signals with the other three signals and is used are average as "Actual Rx Pressure".
0610 NRC RO EXAM 5 RO 295026EA2.01 001lCIA/Tl/G1II/295026EA2.01llRO/SROIBANK Given the following plant conditions:
* Unit 2 is in a transient condition with current conditions as follow Suppression pool level: 13.5 feet - Reactor pressure: 900 psig - Suppression pool temperature: 105°F Which ONE of the following describes the required action?
REFERENCE PROVIDED A'I Operate ALL available Suppression Pool Coolin B. Rapidly depressurize via the Main Turbine Bypass Valve C. Emergency Depressurize the RPV by opening ALL six ADS Valve D. Lower Reactor Pressure to stay within the Safe Area of the Heat Capacity Temperature Limit Curve and maintain cooldown rate below 100°F/h KIA Statement: 295026 Suppression Pool High Water Temp. /5 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 correctly identify an adverse condition related to Suppression Pool High Temperature and then determine the action required to correct the adverse conditio Reference: 2-EOI-2 Flowchart, EOIPM Section O-V-D Page 85 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 solve a problem. This requires mentally using this knowledge and its meaning to resolve the proble NRC Exam Friday, February 29,20083:01 :06 AM 110
0610 NRC RO EXAM REFERENCE PROVIDED: HCTL Curve only Plausibility Analysis: A - correct: 8 - incorrect: ED would not be anticipated under these conditions unless the candidate focuses on Suppression Pool (Torus) water level which is approaching the limit of 11.5 feet for Emergency Depressurization. If this is the case, other actions on SP/L (Suppression Pool I Level leg) take priority over E C - incorrect: No condition has been met requiring Emergency Depressurization at this temperature. It is plausible if the candidate focuses on the SP level, which is approaching the limit of 11.5 feet for Emergency Depressurization. If this is the case, other actions on SP/L take priority over E o - incorrect: It is plausible if the candidate continues down the SPIT leg of EOI-2 and determines that exceeding the HCTL is possible. Since EOI-1 is used to lower pressure and cooldown, and no EOI-1 entry condition has been met, it is unacceptable to assume that exceeding the HCTL is possible.
( Friday, February 29,20083:01 :06 AM 111
2*E,01 ..2.*. .PRIMARY<CONTAINiME;NT*CONTROL
,. * * .': * <'. * :- - ,.' .' ," .".. - * '. ~ , "';' * ".- - * , - " -' ,-' "" '- -. - " *- " ** < - - *** ,. * '-,' ". , * , * :. '. * - ;-' - ' . * 2*1:01*2 gill I:: ... i:;
OU!l'n2 IIIIJIH"WLlfllt1"
.
OlIO _
--,.........................
_ fll>> ..., __ EOI*2 ~ :::~::-:.=.t:'.::~-"
............ , ...
~AVlIIII ==~::~:=."'--
_~_ .... 'MIfl fo ,= JlII'..... "" .... - PRIMARY CONTAINMENT CON BRlJ~ST1ERRY NUCLEll.RPLANT
EXAMINATION REFERENCE . PROVIDED TO CANDIDATE
CURVE 3 HEAT CAPACITY TEMP LIMIT i60+-~ __~__+-~~~~~-+~~__~-+~+-~~~~~~ 150~~--~--+-~--~-----+--~--~----~----~----~ 11.5 12 13 14 i5 16 -rt 18 19 SUPPR PL LVL (FT) ACTION REQUIRED If ABOVE CURVE FOR EXISTING R.:X PRESS
0610 NRC RO EXAM 5 RO 295028EK3.04 OOllelA/Tl G 1I480VLS/B5/295028EK3 .041IRO/SRO/BANK Given the following plant conditions:
* A Loss of Off-site power has occurred in conjunction with a LOCA on Unit * Plant conditions are as follows: - Reactor Water Level (+) 20 inches, steady - Average Drywell Temperature 230°F, rising - Suppression Chamber Pressure 11 psig, rising - Emergency Diesel Generators Tied and loaded to 4 KV Sd Bds - Reactor pressure Remains greater than 800 psig Which ONE of the following describes the final status of Unit 2 Drywell cooling and Reactor Building Closed Cooling Water (RBCCW)?
A." Drywell coolers are operating and RBCCW is availabl Drywell coolers are operating; but, RBCCW is NOTavailabl Drywell coolers must be manually restarted and RBCCW is availabl Drywell coolers must be manually restarted; but, RBCCW is NOT available.
( KIA Statement: 295028 High Drywell Temperature 15 EK3.04 - Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL TEMPERATURE: Increased drywell cooling KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the status of drywell cooling following a transient which results in high drywell temperatur 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 outcom NRC Exam Friday, February 29,20083:01 :06 AM 112
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Whether a Common Accident Signal (CAS) 480V Load Shed has been initiated based on the given condition . The status of RBCCW and OW Blowers based on the answer to Item #1 abov A - correct: B - incorrect: This is plausible because the Orywell Blowers would be operatin However, RBCCW does not receive a trip signal because a 480V Load Shed signal has NOT yet been initiated. RPV level and pressure are too hig C - incorrect: This is plausible because following a 480V Load Shed, the Orywell Blowers, on the accident unit, MUST be manually started and RBCCW would be available. However, a 480V Load Shed signal has NOT yet been initiated. RPV level and pressure are too hig D - incorrect: This is plausible because following a 480V Load Shed, the Orywell Blowers, on the accident unit, MUST be manually started. However, RBCCW does NOT receive a trip signal because a 480V Load Shed signal has NOT yet been initiated.
Friday, February 29, 2008 3:01 :06 AM 113
OPL 171'.072 Revision 11 Page 7 of 30 INSTRUCTOR NOTES x. Lesson Body The 480V Load Shedding Logic System removes selected Obj. V.B.1N. loads from 480V boards which are powered from the 4kV TP-1, 2 Shutdown Boards The load shedding is initiated by an accident signal Obj. V.B.3! V. on Unit 1 or 2 with! diesel generator supplying Obj. V. one 4kV Shutdown Board as its only source of power AND The accident signal is generated in the Core Spray TP-3 System logic Obj. V.B.2.N. Obj. V. Low-low-low reactor water level (-122"/Level 1) OR CASAsignal High drywell pressure (2.45 psig) with low reactor pressure (450 psig) For load shed signal on U1 or U2, the accident is for either unit Unit 3 accident signal won't cause Unit 1 or 2 load shed or vice versa The signal representing "diesel generator supplying TP-4 a 4KV shutdown board" is called "DGVA" For DGVA logic to be satisfied, both conditions must be present: The DG output breaker or the U2 tie breaker to U3 being closed The normal and alternate feeder breaker must be open All Unit 1-2 DGVA contacts are in parallel Any D!G tied to its Shutdown Board with an accident signal present will initiate U 1-2 load shed logiC
BFN Loss of Offsite Power (161 and SOO 0-AOI-S7-1A Unit 0 KV)/Station Blackout Rev. 0071 ( Page 7 of 71 AUTOMATIC ACTIONS (continued) Unit 1/2 480V Load Shed occurs on a loss of offsite power in conjunction with a LOCA signal: One RBCCW pump auto restarts (after 40 seconds on U1 and U2). Drywell Blowers auto restart on non-accident unit (after 40 seconds).
Drywell Blowers with their respective auto restart inhibit switches in the INHIBIT position will not auto restar . Drywell coolers are manually restarted on the accident unit. A Drywell Blower with its auto restart inhibit switch in the INHIBIT position can be manually restarted after a ten minute time dela . SGT TRAINS A & B trip, but will AUTO RESTART in 40 seconds when an initiation signal is presen . Loss of Control Bay Chilled Water Pumps A & B. (may be restarted after 10 minutes with use of bypass switch). Unit 3 480V load shedding occurs as follows: Division I 480V load shedding will occur when an accident signal is present and diesel generator voltage is available on the 4160V shutdown board supplying the 480V shutdown board 3A as follows: RBCCW pump 3A trips Drywell blowers 3A 1 & 3A2 trip After a 40 second time delay, with the control switch in Normal After Start, RBCCW pump 3A restarts After a 40 second time delay, Drywell blowers 3A1 and 3A2 can be manually restarted Drywell blowers 3A3, 3A4 and 3A5 cannot be restarted until the load shed signal is corrected
0610 NRC RO EXAM 54. RO 295030EAl.06 OOllCIA/TlG1I3.5/3.5//295030EAl.06//RO/SROIBANK Given the following plant conditions:
* A LOCA has caused gross fuel failure on Unit * The Site Emergency Director (SED) I SRO has approved implementation of EOI Appendix 18, "Suppression Pool Water Inventory Removal and Makeup."
- The control room crew has just CLOSED the RHR RADWASTE SYS FLUSH VALVE,3-FCV-74-6 * Suppression Pool level is (-)3.5 inches and stead Which ONE of the following describes the next appropriate action(s)?
A. Verify OPEN 3-FCV-73-40, HPCI CST SUCTION VALVE, and OPEN 3-FCV-73-30, HPCI PUMP MIN FLOW VALV B. OPEN 3-FCV-74-63, RHR RADWASTE SYS FLUSH VALVE, and direct Suppression Pool water to Radwaste ONL C. OPEN 3-FCV-74-62, RHR MAIN CNDR FLUSH VALVE, and direct Suppression Pool water to the Main Condenser ONL D~ Inform the SRO that Appendix 18, "Suppression Pool Water Inventory Removal and Makeup" is COMPLETE and that Suppression Pool level is acceptable.
Friday, February 29, 2008 3:01 :06 AM 114
0610 NRC RO EXAM KIA Statement: 295030 Low Suppression Pool Water Level I 5 EA 1.06 - Ability to operate andlor monitor the following as they apply to LOW SUPPRESSION POOL WATER LEVEL: Condensate storage and transfer (make-up to the suppression pool): Plant-Specific KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the effectiveness of actions to control Suppression Pool level using the Condensate storage and transfer syste References: 2-EOI Appendix 18 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 outcom NRC Exam REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Which actions are required based on the given condition NOTE: Each distractor is plausible because they are all actions directed by 2-EOI Appendix 18 "Suppression Pool Water Inventory Removal and Makeup" to control Suppression Pool (Torus) water leve D - correct: A - incorrect: The given Suppression Pool (Torus) water level is sufficiently high enough that additional inventory makeup is NOT necessar B - incorrect: The given Suppression Pool (Torus) water level is sufficiently low enough that additional inventory removal is NOT necessary. However, following a gross fuel failure, rejecting water to Radwaste is more appropriate than to the main condense C - incorrect: The given Suppression Pool (Torus) water level is sufficiently low enough that additional inventory removal is NOTnecessary. In addition, following a gross fuel failure, rejecting water to the main condenser is also NOT appropriate.
Friday, February 29,20083:01 :06 AM 115
3-EOI APPENDIX-18 Rev. 2 Page 1 of 4
.------------------------.
3-EOI APPENDIX-18 SUPPRESSION POOL WATER INVENTORY REMOVAL AND MAKEUP LOCATION: Unit 3 Control Room ATTACHMENTS: None
CAUTION
[NRC/C) Suppression Pool water ~ill be highly radioactive after a LOC Chemical Engineering recommendations are used to determine location to pump contaminated wate [NRC Inspection Report 89-16)
NOTE: All panel operations performed at Control Room Panel 3-9-3 unless otherwise state . IF ..... Suppression Pool Water makeup is required, THEN ... CONTINUE in this procedure at Step . IF ..... Gross fuel failure is suspected, THEN ... OBTAIN SED/SRO permission to pump down Suppression Pool BEFORE continuing in this procedur . IF ..... Directed by SRO, THEN ... REMOVE water from Suppression Pool as follows: DISPATCH personnel to perform the following (Unit 3 RB, El 519 ft, Torus Area) : 1) VERIFY OPEN 3-SHV-074-0786A(B), RHR DR PUMP A(B) DISCH SHUTOFF VALV ) OPEN the following valves:
* 3-SHV-074-0564A(B), RHR DR PUMP A(B) SEAL WTR SPLY ____ * 3-SHV-074-0529A(B), RHR DR PUMP A(B) SHUTOFF VL ) UNLOCK and OPEN 3-SHV-074-0765A(B), RHR DR PUMP A(B) DISC ) NOTIFY Unit Operator that RHR Drain Pump 3A(3B) is lined up to remove water from Suppression Poo ) REMAIN at torus area UNTIL Unit 3 Operator directs starting of RHR Drain Pump 3A(3B) .
3-EOI APPENDIX-18 Rev. 2 Page 2 of 4 (continued from previous page) IF ..... Main Condenser is desired drain path, THEN ... OPEN 3-FCV-74-62, RHR MAIN CNDR FLUSH VALV IF ..... Radwaste is desired drain path, THEN ... PERFORM the following: 1) ESTABLISH communications with Radwast ) OPEN 3-FCV-74-63, RHR RADWASTE SYS FLUSH VALV NOTIFY personnel in Unit 3 EB, El 519 ft, Torus Area to start RHR Drain Pump 3A(3B) . THROTTLE 3-FCV-74-108, RHR DR PUMP 3A/B DISCH HDR VALVE, as necessary.
4. WHEN Suppression Pool level reaches -5.5 in., THEN SECURE RHR Drain System as follows: DISPATCH personnel to STOP the Drain System as follows (Unit 3 RB, El 519 ft, Torus Area) : 1) STOP RHR Drain Pump 3A (3B).
2) CLOSE the following valves:
* 3-SHV-074-0564A(B) , RHR DR PUMP A(B) SEAL WTR SPLY * 3-SHV-074-0529A(B) , RHR DR PUMP A(B) SHUTOFF VL ) CLOSE and LOCK 3-SHV-074-0765A(B) , RHR DR PUMP A(B) DISC CLOSE 3~FCV-74-108, RHR DR PUMP 3A/B DISCH HDR VALV VERIFY CLOSED 3-FCV-74-62, RHR MAIN CNDR FLUSH VALV VERIFY CLOSED 3-FCV-74-63, RHR RADWASTE SYS FLUSH VALV WHEN ... Suppression Pool level can be maintained between -1 in. and -5.5 in.,
THEN ... EXIT this procedur EOI APPENOIX-18 Rev. 2 Pa e 3 of 4 IF ..... Directed by SRO to Emergency Makeup to the Suppression Pool from Standby Coolant, THEN ... CONTINUE in this procedure at Step 9. IF ..... Directed by SRO to add water to suppression pool, THEN .. . MAKEUP water to Suppression Pool as follows: VERIFY OPEN 3-FCV-73-40, HPCI CST SUCTION VALV OPEN 3-FCV-73-30, HPCI PUMP MIN FLOW VALV IF ..... HPCI is NOT available for Suppression Pool makeup, THEN .. . MAKEUP water to Suppression Pool using RCIC as follows: 1) VERIFY OPEN 3-FCV-71-19, RCIC CST SUCTION VALV ) OPEN 3-FCV-71-34, RCIC PUMP MIN FLOW VALV IF ..... 3-FCV-71-34, RCIC PUMP MIN FLOW VALVE, CANNOT be opened from control room, THEN . .. DISPATCH personnel to 250V DC RMOV Board 3B, Compartment 50, to perform the following: 1) PLACE 3-XS-071-0034, RCIC PUMP MIN FLOW VALVE EMER TRANS SWITCH, to EMER ) OPEN 3-FCV-71-34, RCIC PUMP MIN FLOW VALVE.
7. WHEN ... Suppression Pool level reaches -5.5 in., THEN . .. VERIFY CLOSED the following valves:
* 3-FCV-73-30, HPCI PUMP MIN FLOW VALVE * 3-FCV-71-34, RCIC PUMP MIN FLOW VALVE.
8. DISPATCH personnel to 250V DC RMOV Board 3B, Compartment 50, to VERIFY 3-XS-071-0034, RCIC PUMP MIN FLOW VALVE EMER TRANS SWITCH, in NORMA NRC RO EXAM 5 RO 29503IG2.4.6 001/C/A/TlGlIC1II29503IG2.4.6//RO/SROIBANK Given the following plant conditions: (
* Unit 2 was operating at 98% power when an automatic scram occurs due to a Group I Isolatio * ALL control rods fully insert as reactor water level immediately drops below Level * BOTH Recirc Pumps tri * HPCI automatically initiates but immediately isolates due to a ruptured inner turbine exhaust rupture diaphrag * RCIC is currently tagged out of service to repair an oil lea * ALL other systems are operabl * EOI-1, RPV Control, is entere * Pressure control was established with Safety Relief Valves (SRVs).
The remaining high pressure injection systems are unable to maintain reactor water level which is currently at (-)150 inches and lowerin Which ONE of the following contingency procedures would be appropriate to execute? A.'" C1, "Alternate Level Control." C2, "Emergency RPV Depressurization." C4, "RPV Flooding." C5, "Level/Power Control."
KIA Statement: 295031 Reactor Low Water Level 12 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 Emergency Procedure used to mitigate a low reactor water level conditio References: 2-EOI-1, EOIPM Sections O-V-C and O-V-G 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 outcom NRC Exam Friday, February 29,20083:01 :06 AM 116
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Whether the given conditions are indicative of a loss of HP injectio . Based on Item #1 above, which EOI Contingency is appropriate to mitigate that conditio A - correct: B - incorrect: This is plausible since Emergency Depressurization (ED) will eventually become necessary following the initial actions of EOI-C1. However, additional actions are required before EOI-C2 is appropriat C - incorrect: This is plausible since Drywell temperature may be high enough, following Emergency Depressurization, to create environmental conditions where RPV level instruments become inaccurate / unavailable. However, additional actions are required before EOI-C4 is appropriat D - incorrect: This is plausible since the ONLY given condition which contradicts the use of EOI-C5 is the current rod pattern. However, with ALL rods inserted, EOI-C5 is NOT appropriate.
Friday, February 29, 2008 3:01 :06 AM 117
-o
~
I IIII i ii
~i ,I !
II
!
j
! ~;
II Ii N ..o*
~.
III
..-o*
N* III
0610 NRC RO EXAM 5 RO 295037EK2.11 001lCIA/TlGIIRMCSII295037EK2.1111RO/SROIBANK A hydraulic ATWS has occurred on Unit 2 and the Operator At-The-Controls is inserting control rods in accordance with the following:
* EOI Appendix 1 D, "Insert Control Rods Using Reactor Manual Control System," * EOI Appendix 1F, "Manual Scram," and * EOI Appendix 2, "Defeating ARI Logic Trips."
During execution of these appendices, _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ A." ALL potential Control Rod Insert Block signals are bypasse Rod Drift indication is received as soon as rod motion begin Stabilizing Valves are OPEN to provide increased drive water pressur ALL Reactor Manual Control System (RMCS) timer functions are bypassed except for the Settle Bus Time KIA Statement: 295037 SCRAM Condition Present and Power Above APRM Downscale or Unknown/1 EK2.11 - Knowledge of the interrelations between SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN and the following: RMCS: Plant-Specific KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the status of the RMCS while executing procedures to mitigate an ATWS conditio References: 2-EOI Appendicies 1D, 1F, and 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 outcom NRC Exam Friday, February 29, 2008 3:01 :06 AM 118
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. What affect the actions performed by EOI Appendix implementation have on the RMCS syste . What affect the RMCS manipulations required by implementation of the EOI Appendices have on plant indication A - correct: B - incorrect: This is plausible since a Rod Drift indication will occur for each inserted control rod. However, the indication does NOT occur until the rod is fully inserted and the CRD NOTCH OVERRIDE switch is release C - incorrect: This is plausible because CRD Stabilizing Valves do have an effect on drive water pressure, but the efffect is to prevent oscillations while moving control rods, NOT increase pressur D - incorrect: This is plausible since RMCS timers are bypassed by using the CRD NOTCH OVERRIDE switch in accordance with EOI Appendix 1D. However, the Settle Bus Timer is also bypassed.
Friday, February 29,20083:01 :06 AM 119
2-EOI APPENDIX-1D Rev. 6 Page 1 of 3
.------------------------.
2-EOI APPENDIX-1D INSERT CONTROL RODS USING REACTOR MANUAL CONTROL SYSTEM LOCATION: Unit 2 Control Room, Panel 9-5 'I __A_T_T_A_C_H_M_E_N_T_S_:__ _._T_O_o_l_s_a_n_d_E_q_U_l_* Core Position Map p_m--"'-e_n_t_ _ _ _ _ _ _--:-_ _ _ _ _ _-.:...~~
(,/)
NOTE: This EOI Appendix may be executed concurrently with EOI Appendix 1A or IB at SRO's discretion when time and manpower permi I VERIFY at least one CRD pump in servic NOTE: Closing 2-85-586, CHARGING WATER ISOL, valve may reduce the effectiveness of EOI Appendix 1A or l . IF ..... Reactor Scram or ARI CANNOT be reset, THEN ... DISPATCH personnel to close 2-SHV-85-586, CHARGING WATER SHUTOFF (RB NE, El 565 ft). VERIFY REACTOR MODE SWITCH in SHUTDOW . BYPAS~ Rod Worth Minimize . REFER TO Attachment 2 and INSERT control rods in the area of highest power as follows: SELECT control ro PLACE CRD NOTCH OVERRIDE switch in EMERG ROD IN position UNTIL control rod is NOT moving inwar REPEAT Steps 5.a and 5.b for each control rod to be inserte NOTE: A ladder may be required to perform the following step. REFER TO Tools and Equipment, Attachment IF necessary, an alternate ladder is available a the HCU Modules, EAST and West banks. It is stored by the CRD Charging Car . WHEN ... NO further control rod movement is possible or desired, THEN ... DISPATCH personnel to verify open 2-SHV-85-586, CHARGING WATER SHUTOFF (RB NE, El 565 ft) . END OF TEXT
2-EOI APPENDIX-IF Rev. 4
..----------------------..
2-EOI APPENDIX-1F Page 1 of 7 MANUAL SCRAM l j LOCATION: Unit 2 Control Room I ATTACHMENTS: 1. Tools and Equipment I 2. Panel 2-9-15, Rear 3. Panel 2-9-17, Rear
!
II VERIFY Reactor Scram and ARI rese IF ..... ARI CANNOT be reset, THEN ... EXECUTE EOr Appendix 2 concurrently with Step 1.b of this procedur I IF ..... Reactor Scram CANNOT be reset, THEN ... DISPATCH personnel to Unit 2 Auxiliary Instrmnent Room to defeat ALL RPS logic trips as follows: 1) REFER to Attachment 1 and OBTAIN four 3-ft banana jack jumpers from EOI Equipment Storage Bo ) REFER to Attachment 2 and JUMPER the following relay terminals In Panel 2-9-15, Rear: a) Relay SA-K10A (DQ) Terminal 2 to Relay I I 5A-K12E (ED) Terminal 4, Bay b) Relay 5A-K10C (AT) Terminal 2 to Relay 5A-K12G (BH) Terminal 4, Bay ) REFER to Attachment 3 and JUMPER the following relay terminals in Panel 2-9-17, Rear: aJ Relay 5A-KIOB (DQ) Terminal 2 to Relay 5A-K12F (ED) Terminal 4, Bay . b) Relay 5A-I{10D (AT) Terminal 2 to Relay 5A-K12H (BH) Terminal 4, Bay WHEN ... RPS Logic has been defeated,
-!
I THEN ... RESET Reactor Scra I VERIFY OPEN Scram Discharge Volume vent and drain valves.
I II I~I========================================~.I I,
0610 NRC RO EXAM 5 RO 295038EK1.01 001lMEM/TlG1INEWIl295038EK1.01llRO/SRO/MODIFIED 11117/07 Given the following plant conditions:
* Unit 2 has experienced a LOCA with a loss of Primary Containmen * You have volunteered for a team, dispatched from the Operations Support Center (OSC), that will enter the Reactor Building and attempt to energize '2D' 480V RMOV boar * Due to environmental and radiological conditions present in the Reactor Building, Radiation Protection personnel provide each team member with a Sodium Chloride (NaCI) and Potassium Iodide (KI) tablet during the briefin Which ONE of the following describes the benefit of ingesting Potassium Iodide prior to the Reactor Building entry?
Potassium Iodide will reduce the
-------------------------------------
A. risk of dehydration and heat stres B. absorption of radioactive Iodine in the lung C~ absorption of radioactive Iodine by the thyroi D. absorption of radioactive Potassium in the blood strea KIA Statement: 295038 High Off-site Release Rate I 9 EK1.01 - Knowledge of the operational implications of the following concepts as they apply to HIGH OFF-SITE RELEASE RATE: Biological effects of radioisotope ingestion KIA Justification: This question satisfies the KIA statement by requiring the candidate to correctly identify the pathway and adverse effect of iodine ingestio Reference: EPIP -14 Revision 18, page 4 Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam MODIFIED FROM 295038EK1.01 #1 Friday, February 29, 2008 3:01 :06 AM 120
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: C - correct: A - incorrect. The sodium chloride tablets would be used for this purpose. It is plausible if the candidate is unsure of the purpose of KI tablet B - incorrect: Only the thyroid is the organ at risk, but it is plausible if the candidate assumes that airborne ingestion is limited to absorption by the lung D - incorrect: Iodine is the element that is absorbed. Potassium becomes a plausible answer due to recent media coverage regarding health risks related to low potassium levels in the blood stream.
( Friday, February 29, 2008 3:01 :06 AM 121
BROWNS FERRY RADIOLOGICAL CONTROL PROCEDURES EPIP-14 Issuing Potassium Iodide (KI) 3. If the TSC RP Manager has reason to believe that a person's projected cumulative dose to the thyroid from inhalation of radioactive iodine might exceed 10 rems (see Appendix A), the exposed person should be started immediately on a dose regimen of KI. This decision shall be immediately communicated to the SE .6. If the TSC is not staffed or the RP Manager position has not been filled, then the senior onsite RP Supervisor has the authority to issue KI utilizing the bases described in step 3. .6.1.2 The initial dose of KI should be not delayed since thyroid blockage requires 30 to 60 minutes. Anyone authorized to initiate KI shall be familiar with the Food and Drug Administration (FDA) patient package insert and be sure that each recipient is similarly informe .6.1.3 Prior to issuing KI to an individual, the person should be asked if he/she is allergic to iodine. If the person indicates a possible sensitivity to iodine they should not be issued K . KI is stored in the plant RP supply cage and the REP Van instrument kit . RP normally will not dispense a container or package of KI to TVA Personnel involved in activities to support a radiological emergency. RP will however dispense a single individual dose of KI to team members dispatched from the OS . Follow the dosage outlined on the FDA patient package insert (Appendix B). A copy of the FDA approved patient package insert shall accompany the issuance of KI. If KI is distributed in individual doses then verbal instructions of the significant information on the patient package insert by a knowledgeable individual is sufficien . Complete the KI Issue Report (Appendix C) or document on an RWP time sheet as appropriate for issuance of KI. If the RWP time sheet is used to document distribution of the KI, note the time of KI distribution on the back of the time shee PAGE 4 OF 9 REVISION 0018
0610 NRC RO EXAM 5 RO 600000AAl.08 001IMEMlT1Gl/RSWI1600000AAl.08//RO/SROINEW 11120107 RMS Which ONE of the following describes the appropriate fire extinguishing agent for the specific class of fire? Water used on Class 'B' fire B." Low pressure C02 used on Class 'c' fire Dry Chemical (PKP) used on Class 'c' fire Aqueous Film Forming Foam (AFFF) used on Class 'A' fire KIA Statement: 600000 Plant Fire On-site I 8 AA 1.08 - Ability to operate and lor monitor the following as they apply to PLANT FIRE ON SITE: Fire fighting equipment used on each class of fire KIA Justification: This question satisfies the KIA statement by requiring the candidate to identify the correct fire fighting agent for a specific class of fir References: TVA Safety Manual Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam Friday, February 29, 2008 3:01 :06 AM 122
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Which flammable material is of concern based on Fire Class A, 8 and . Which extinguishing agent is appropriate for each class of fir . Which extinguishing agent is inappropriate for a given class of fir B - correct: A - incorrect: Class "8" fires are flammable liquids. Using water could cause serious damage by allowing the liquid to splatter and sprea C - incorrect: Dry chemical agents are extremely corrosive to electrical components and insulation typical of Class "8" electrical fire D - incorrect: AFFF is designed as a flooding and diluting agent for Class "8" flammable liquid fires. Application on a Class "A" fire is not effective in extinguishing flammable materials such as wood and paper.
Friday, February 29, 2008 3:01 :06 AM 123
0610 NRC RO EXAM 59. RO 295009AK2.01 001lC/A/TlG2IPR.INSTRJ131295009AK2.01l9619/RO/SROINEW 10116/07 Given the following Unit 1 plant conditions: (
* Due to multiple high pressure injection system failures, 1-EOI-C1, "Alternate Level Control" has been entere * RHR Pump '1A' is running and lined up for LPCI injectio * Core Spray Pumps '1 B' and '10' are running and lined up for injectio * Drywell Temperature is 240 of and rising slowl Which ONE of the following conditions describes the appropriate point where Emergency Depressurization may be performed in accordance with 1-EOI-C1, "Alternate Level Control?"
Post Accident Flooding Range Level Instrument, 3-Ll-3-52, is reading (1) with reactor pressure at (2) REFERENCES PROVIDED ( (-)205 inches; 350 psig ( (-)210 inches; 500 psig (-)220 inches; 900 psig D..... (-)225 inches; 800 psig Friday, February 29, 2008 3:01 :06 AM 124
0610 NRC RO EXAM KIA Statement: 295009 Low Reactor Water Level I 2 AK2.01 - Knowledge of the interrelations between LOW REACTOR WATER LEVEL and the following: Reactor water level indication KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine actual reactor water level under conditions of low reactor water leve References: 1-EOI-C1 Flowchart, PIP-95-64 Rev 12 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 outcom NRC Exam REFERENCE PROVIDED: 1-EOI-C1 Flowchart, PIP-95-64 Rev. 12 Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Recognize the requirement that RPV level must be less than -162 inches before Emergency Depressurization is appropriat . Recognize that the indicated RPV level must be corrected for pressure using PIP-95-6 . Recognize that two or more injection systems must be lined up with pumps running to meet the requirement to Emergency Depressuriz . Recognize that only one RHR pump is required to qualify as an injection subsystem since each RHR pump is rated for 100% capacit NOTE: Each distractor is plausible because the conditions specified are possible given the current plant condition correct: A - incorrect: Level is 5 inches too high or pressure is 100 psig too hig incorrect: Level is -4 inches too high or pressure is 240 psig too hig C - incorrect: Level is -4 inches too high or pressure is 100 psig too high.
Friday, February 29, 2008 3:01 :06 AM 125
3-U-3-52 & 62 CORRECTION CURVES i i !)) II t r 1 ! iii! TAF I i-175"
,
I-l W !'c !!
>
W i f'fc, I: I I ~J I : I I r ! I iJl-l -200" I' !
, ! '1°(", i!' I <+' ~ ! Ii [[ i i iI i i'
o 1'1! I I",'
~J I
, i I I, t" ',i I, i W 1-r--l-++-t-+-+++++H-++-i-+'ii"clcr_-++-t'I"H-tI'-'-,ltLLJ+-H-t-++HH-++t-HI-+j-+++rHl'-++-f--Hi1-+'+1 ACTUAL I-
<<
<-)
, ; 1\ LEVEL-225 (\ ! [
I I o I" i I I z I' T'kl
, !
II 1 ' I' ,
-~
l' ! 1(- Ji
,~: ~ I II ~ -250" ! i't-- I j I i I I 1]-1 II ! iI i i \
iI I I
; ,i ! !
i i
[ I'
i , i
! I I
I ,I
./
" a i! I I . i i[ I
! !I I I !
./
" -215" I-268" o 100 200 300 400 500 600 700 800 900 1000 1100 REACTOR PRESSURE (PSIG) PIP-95-64 REV. 12
OPL 171.003 Revision 17 Page 22 of 54 INSTRUCTOR NOTES Since no trips or alarms are associated with this range, this level signal is not directed through the Analog Trip Syste (d) One MCR indicator on Panel 9-3 monitors this range of level indication.
(4) Post-accident Flood Range (a) -268" to +32" range covering active core area and overlapping the lower portion of the Normal Control Rang (b) Referenced to instrument zero (c) Intended for use only under accident conditions with reactor at 0 psig and recirculation pumps trippe (d) Variable leg tap is from diffuser of jet pumps 1 and 6 (or 11 and 16).
(e) Per Safety Analysis on water level instruments the conclusion Injecting with RHR is that the accident range U-3-52 and 62 instruments adequately indicate (Accident Range) water level--provided they are Technical Support corrected for off-calibration letter dated 9/13/95 conditions of RPV pressure (See LP Folder) utilizing the operator aid on Panel Use Conservative 9-3 for level correctio Decision Making Obj. V.B.1 (f) An interlock associated with this Obj. V.B.1 range will prevent using the RHR System for containment de pressurization when it is needed to flood the core regio (g) The -68" to -168" portion of this range is recorded in the MCR on Unit 3 Recorder 2-U-3-62 Recorder and two displays a scale of indicators monitor the full range +32" to -268" of these instrument EXAMINATION REFERENCE PROVIDED TO CANDIDATE
~ I UI I- ...o III:: I-Z o u ... II > II U II l-CC Z III:: II ...cc I- ...I U ...I
~
I U
( 3-Ll-3-52 & 62 CORRECTION CURVES-150" TAF-175"
-1 ill >
ill-1 -200" o ill
'- << ()
o -225 z-250"
; -268" o 100 200 300 400 500 600 700 800 900 1000 1100 REACTOR PRESSURE (PSIG) PIP-95-64 REV. 12
0610 NRC RO EXAM 60. RO 295012G2.2.22 OOllC/A/TlG2/64/121295012G2.2.221IRO/SRO/BANK Given the following plant conditions:
* You are the oncoming Unit 3 Unit Superviso * During turnover, the onshift Unit Supervisor informs you that two (2) Drywell Coolers had been secured during his shift while performing ground isolation on '3C' 480V RMOV boar * Drywell Average Temperature is 152°F and stabl Which ONE of the following describes the appropriate condition and required action?
A'!' Exceeded 3-SR-2, "Instrument Checks and Observations", Drywell temperature limit. Address Tech Spec section B. Exceeded the normal operating Drywell temperature limit. Drywell temperature must be logged hourly until below the limi C. Exceeded the normal operating Drywell temperature limit. Restore Drywell average air temperature below the limit in 24 hour D. Exceeded 3-EOI-2, "Primary Containment Control" entry condition. Enter and execute 3-EOI-2, Primary Containment Contro KIA Statement: 295012 High Drywell Temperature I 5 2.2.22 - Equipment Control - Knowledge of limiting conditions for operations and safety limit KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine that Technical Specification limits have been exceede References: Unit 3 Tech Specs Section 3.6. 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 outcom NRC Exam Friday, February 29,20083:01 :06 AM 126
0610 NRC RO EXAM REFERENCE PROVIDED: 1-EOI-C1 Flowchart, PIP-95-64 Rev 12.
( Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. The appropriate entry condition for U3 Tech Spec Section 3.6. . The appropriate entry condition for 3-EOI-2, Primary Containment Contro . The appropriate action based on the given conditio A - correct: B - incorrect: This is plausible because the Tech Spec limt was exceeded. However, the required action is to restore the Orywell Temperature within the limit in 8 hour There is NO requirement for hourly logging of OW temperatur C - incorrect: This is plausible because the Tech Spec limt was exceeded. However, the required action is to restore the Orywell Temperature within the limit in 8 hour The 24 hour limit is based on performing the surveillance on Orywell Temperatur D - incorrect: This is plausible because the entry condition for 3-EOI-2 is only 8 of above the given temperature. However, the entry condition has NOT been met and OW temperature was reported as "stable".
( Friday, February 29,20083:01 :06 AM 127
Drywell Air Temperature 3.6.1.4 3.6 CONTAINMENT SYSTEMS 3.6.1.4 Drywell Air Temperature LCO 3.6. Drywell average air temperature shall be::::; 150°F.
APPLICABILITY: MODES 1, 2, and 3.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell average air Restore drywell average 8 hours temperature not within air temperature to within limi limi B. Required Action and Be in MODE hours associated Completion Time not me AND Be in MODE hours BFN-UNIT 3 3.6-17 Amendment No. 212
Drywell Air Temperature 3.6.1.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.1. Verify dryweil average air temperature is 24 hours within limit.
BFN-UNIT 3 3.6-18 Amendment No. 212
0610 NRC RO EXAM 61. RO 295015AKl.02 001lMEM/TlG2/BASISI1295015AKl.0211IBANK Which ONE of the following describes why the Automatic Depressurization System (ADS) is inhibited once Standby Liquid Control injection has begun in accordance with EOI-1, "RPV Control" path RC/Q? The operator can control pressure better than an automatic system like AD ADS actuation would impose a severe pressure and temperature transient on the reactor vesse C." Severe core damage from a large power excursion could result, if Imv pressure
< systems automatically injected on dQpress'lrizatio If only steam driven high pressure injection systems are available an ADS actuation could lead to a loss of adequate core coolin KIA Statement:
295015 Incomplete SCRAM / 1 AK1.02 - Knowledge of the operational implications of the following concepts as they apply to INCOMPLETE SCRAM: (CFR 41.8 to 41.10) Cooldown effects on reactor power KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the effect of a significant cooldown when an incomplete scram has occurre References: Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam Friday, February 29, 2008 3:01 :06 AM 128
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. The basis for inhibiting ADS under the specific conditions of boron injectio NOTE: Each of the three distractors are plausible based on their relationship to the bases for inhibiting ADS under circumstances OTHER than boron injectio Specifically, Alternate RPV Level Control actions. Refer to the attached excerpt from EOIPM Section O-V- C - correct: A - incorrect: This statement is true, but is NOT addressed in the basis for boron injectio B - incorrect: This applies whenever ADS actuates, but is ONLY the precursor to the issue related to boron injectio D - incorrect: This statement applies particularly to a low RPV level condition.
Friday, February 29,20083:01 :06 AM 129
EOI PROGRAM MANUAL EOI-1 RPV CONTROL BASES J SECTION o-v-c
'~> ~ :!,,;' ~ " " S'fEP: ~""
R.El~-14 ana R.(;1~-15, " ~
,
EX(CUTE RC/Q-12 />NO RC/Q-22 CONCVRRENTtY Re/D-II NO L oil ( L RC/Q-13
*
L RC/O-14 INHIBll NJS L RC/Q- 15 VERIFY RWCU SYSlEM ISOLATION L, WHU EXECUTING THE FOLLOWING STEPS IHEl'1 SlC TMlK LVl DROPS TO <A ..3>. TRIP THE SlC PUMPS L RC/Q- 17 034 ( SECTION O-V-C PAGE 116 OF 127 REVISION 1
*
EOI-1, RPV CONTROL BASES EOI PROGRAM MANUAL SECTION O-V-C I DISCUSSION: STEP RC/Q-14 and RC/Q-15 I The RC/Q-14 action step directs the operator to manually initiate the SLC System. Because this step is prioritized
\vith the Ininiature before decision step RC/Q-12 symbol, this action should be performed before suppression pool temperature reaches <A.64>, Boron Injection Initiation Temperature. EOl Appendix 3A provides step-by-step guidance for manual initiation of the SLC System. Boron in solution absorbs neutrons, providing negative reactivity to achieve reactor subcriticality, since the reactor is not yet subcritical on control rod insertion alon The RC/Q-15 action step directs the operator to defeat automatic ADS function by placing the ADS inhibit switches in the inhibit position. Because this step is prioritized with the miniature before decision step RC/Q*l2 symbol, this action should be performed before suppression pool temperature reaches<A.64>, Boron Injection Initiation Temperatur ~~:'if}j,t~~tt9;l}.ma~c'resllit'icJi*.the.cinjecti~ri,Qfll:\f~e~Ocuntsofrela.ti"Velrccold,unbor~tedw!ltgf:r~9Jo~,c pre~sur~
iriJectiohsys:tems.. c' With the reactor still critical or sub critical on boron, the post~tY"':;:*.r~1l'~t1;Yfty;~(id;it.ioil(iU~to boron diJ~1ion0~4t~9;p~r~tpr~,red~~~?~JrOIninjection of cold water mayrestllt in a ~eilctor power excursion large enoti,ghto>cause substahtililcoteCllrmlfge.' Defeating ADS is, therefore, appropriate whenever boron injection is required. If emergency depressurization of the RPV is subsequently required, explicit direction is provided in the
. appropriate EOr. Therefore, the ability to maintain automatic initiation capability of ADS is not required.
- '
( REVISION 1 PAGE 117 OF 127 SECTION O-V-C
EOI PROGRAM MANUAL C1 , ALTERNATE LEVEL CONTROL BASES SECTION O-V-G RC/L-12 I L y r IL
*
INHIBIT ADS C I,,' ~----
--,~,
WHILE EXECUTING THIS PROCEDURE:
~.--~- I.J:i£l' ~--.- , ~ .... -
ALL CONTROL RODS ARE tlOI [XIT HilS PROC[DURE AND INSUnED 10 OR BFYOND ENTER C5. LEVEL/POWER CONlIWL POSITION <A. 70>. RPV WATFR l Vl CANNQI lXIT THIS PROCEDURF AND BE DETERMINE ENTER C4, Rf'II FLOODING RPV WATER LVL IS RISIN EXIT THIS PROCEDURE AND LNffR [01-1. RPV CONTROL, AT STEP RC/l-l CI-2 SECTION O-V-G PAGE 8 OF 50 REVISION 0
*
C1, ALTERNATE LEVEL CONTROL BASES EOI PROGRAM MANUAL SECTION O-V-G . I DISCUSSION: STEP Cl-l I This action step directs the operator to defeat automatic ADS function. An ADS actuation with the RPV at pressure imposes a severe thermal transient on the RPV and may significantly complicate efforts to restore and maintain RPV water level as specified in this procedure .
.
Because ADS initiation logic receives limited input signals, a variety of plant conditions may exist where automatic depressurization of the RPV is not appropriate. In certain cases (e.g., RCIC available but LPCVCS injection valves closed and control power for their operation not available) ADS actuation may directly lead to loss of adequate core cooling and core damage, conditions that might otherwise have been avoided. Further, conditions assumed in the design of ADS actuation logic (e.g., no operator action for ten minutes) do not exist when actions specified in this procedure are being carried ou Finally, an operator can draw on much more plant information than is available to ADS logic (e.g., equipment out of service for maintenance, operating experience with certain systems, probability of restoration of offsite power, etc.) and thus can better judge, based on logic specified in this procedure, when and how to depressurize the RPV. For all of these reasons, it is appropriate to prevent automatic initiation of ADS as specified.
( REVISION 0 PAGE 9 OF 50 SECTION O-V-G
0610 NRC RO EXAM 62. RO 295020AK3.0S OOllMEM/EOIIBASISII295020AK3.0S///BANK Unit 2 was at 100% rated power when a spurious Group I Isolation occurred. The pressure transient caused a small-break tOCA to occur inside the Drywel Which ONE of the following describes the basis for actions before AND after reaching 12 psig Suppression Chamber Pressure? Drywell sprays must be initiated prior to this pressure to prevent opening the Suppression Chamber to Reactor Building vacuum breakers and de-inerting the containmen B.oI Above this pressure indicates that almost all of the nitrogen and other non-condensible gases in the drywell have been transferred to the torus and chugging is possibl Above this pressure indicates that almost all of the nitrogen and other non-condensible gases in the torus have been transferred to the drywell air space and Suppression Chamber Sprays will be ineffectiv Above this pressure indicates that almost all of the nitrogen and other non-condensible gases in the drywell have been transferred to the torus so initiating Drywell Sprays will not result in containment failur KIA Statement: 295020 Inadvertent Cont. Isolation /5 & 7 AK3.08 - Knowledge of the reasons for the following responses as they apply to INADVERTENT CONTAINMENT ISOLATION: Suppression chamber pressure response KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the effect on Suppression Chamber pressure due to an inadvertent containment isolation and the basis for that respons References: EOIPM Section O-V-D Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam Friday, February 29, 2008 3:01 :06 AM 130
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. The basis for the Pressure Supression Pressure Limit of 12 psig Suppression Chamber pressur B - correct: A - incorrect: This is plausible because initiation of OW sprays at high SC pressure could reduce pressure low enough to open the Suppression Chamber to Reactor Building Vacuum Breakers. However, this is part of the bases for the Orywell Spray Initiation Pressure Limit Curve # C - incorrect: This is plausible if the LOCA occurred inside the Suppression Chamber and NOT the Orywell as given in the ste D - incorrect: This is plausible because initiating SC sprays with high temperature non-condensible gases in the SC will result in evaporative cooling and a rapid pressure drop. However, the SC to OW vacuum relief system is capable of compensating for this pressure drop. This is also part of the bases for the Orywell Spray Initiation Pressure Limit Curve #5.
( Friday, February 29, 2008 3:01 :07 AM 131
EOI PROGRAM MANUAL EOI-2, PRIMARY CONTAINMENT CONTROL BASES SECTION O-V-D NO L IS SUPPR PL LVl BELOW <A.B> PC/P-4 YES L o IMTLAT[ SUPPR CHMBR SPRAYS USING 00l.Y RHR PUMPS WI REQI)RED TO ASSURE ADEQUATE CORE COOLING BY CONllNUOVS INJ (l>Ppx 17C) L PC/P-s PC/P-6 SUPPR CHMBR PRESS EXCEEDS <A.65> CONTINUE IN THIS PROCEDURE L
*
IS SUPPR Pl LVl NO L BELOW <A.5> PC/P-l YES L TO CURVE 5 (Refer to EOI Program ManlJo~ "- Section IV, Appendix A, Curves and Tobles Used in the EOls) ARE DW TEMP AND OW PRESS WI'HIN THE NO L SAFE AREA or CURVE 5 PC/P-8 YES L TO PC/P-l1 SECTION O-V-D PAGE 42 OF 244 REVISION 0 *
EOI-2, PRIMARY CONTAINMENT CONTROL BASES EOI PROGRAM MANUAL SECTION O-V-O I DISCUSSION: STEP PCIP-6 I This contingent action step requires the operator to wait until the stated condition has been met before continuing in EOI-2. Performance of subsequent actions in this section ofEOI-2 will not be performed until suppression chamber pressure exceeds Suppression Chamber Spray Initiation Pressur Engineering calculations have detennined that if suppression chamber pressure exceeds <A.65>, Suppression Chamber Spray Initiation Pressure, there is no assurance that chugging will be prevented at downcomer openings of the drywell vents. This value is rounded off in the EO! to use the closest, most conservative value that can be accurately determined on available instrumentatio Suppression Chamber Spray Initiation Pressure is defmed to be the lowest suppression chamber pressure that can occur when 95% of noncondens abies in the drywell have been transferred to airspace of the suppression chamber. Scale model tests have demonstrated that chugging will not occur so long as the drywell atmosphere contains at least 1% noncondensables. To prevent the occurrence of conditions under which chugging may happen, Suppression Chamber Spray Initiation Pressure is conservatively defined by specifying 5% noncondensables .
- Chugging is the cyclic condensation of steam at downcomer openings of the drywell vent Chugging occurs when steam bubbles collapse at the exit of downcomers. The rush of water that fills the void (some of which is drawn up into the downcomer pipe) induces a severe stress at the junction of the downcomer and vent header. Repeated application of this stress can cause these joints to experience fatigue failUre (cracks), thereby creating a pathway that bypasses the pressure suppression function of primary containment. Subsequent steam that discharges through downcomers would then exit through the fatigued cracks, and directly pressurize suppression chamber air space, rather than discharging to and condensing in the suppression poo Although operation of suppression chamber sprays by itself will not prevent chugging, the requirement to wait to initiate drywell sprays until reaching Suppression Chamber Spray Initiation Pressure assures that suppression chamber spray operation is attempted before operation of drywell sprays. Therefore, actions to initiate drywell sprays need to be directed only if suppression chamber sprays were unable to reduce primary containment pressure or they could not be initiated .
- REVISION 0 PAGE 43 OF 244 SECTION O-V-D
EOI-2, PRIMARY CONTAINMENT CONTROL BASES EOI PROGRAM MANUAL SECTION O-V-D
- I DISCUSSION: STEP PC/P-8 This decision step has the operator evaluate the present status of drywell pressure and drywell temperature to determine if conditions are favorable for dryweU spray operatio I Drywell spray operation reduces drywell pressure and temperature through the combined effects of evaporative and convective cooling. During evaporative cooling, water spray undergoes a change of state, liquid to vapor, whereas convective cooling involves no change of stat Evaporative cooling occurs when water is sprayed into a superheated atmosphere. Water at the surface of each droplet is heated and flashes to steam, absorbing heat energy from the drywell atmosphere until the atmosphere reaches saturated conditions. In the dryweU, with a typical drywell spray flowrate, the evaporative cooling process results in an immediate, rapid, large reduction in pressure. This pressure reduction occurs at a rate much faster than can be compensated for by the primary containment vacuum relief system. Unrestricted operation of drywell sprays could cause an excessive negative differential pressure to occur between the drywell and suppression chamber, large enough to cause a loss of primary containment integrit Convective cooling occurs when water is sprayed into a saturated atmosphere. Sprayed water droplets absorb heat from the surrounding atmosphere through convective heat transfer (sensible heat from the atmosphere is transferred to the water droplets). This effect reduces drywell ambient temperature and pressure until equilibrium conditions are established. The convective cooling process occurs at a rate much slower than the evaporative cooling process. An operator can effectively control the magnitude of a containment temperature/pressure reduction from convective cooling by terminating operation of drywell spray Considering the pressure drop concerns described above, engineering calculations have determined that primary containment integrity is assured when drywell sprays are operated in the safe area of Drywell Spray Initiation Limit Curve (Curve 5). DryweU Spray Initiation Limit is defined to be the highest drywell temperature at which initiatioIi of drywell sprays will not result in an evaporative cooling pressure drop to below either: 1) drywell-below-suppression chamber differential pressure capability, or 2) high drywell pressure scram setpoin If drywell temperature and pressure are within the safe area of Curve 5, the operator continues at Step PCiP- If dryweU temperature and pressure are not within the safe area of Curve 5, then drywell spray operation is not permitted, and the operator is directed to Step PCiP-ll .
- REVISION 0 PAGE 47 OF 244 SECTION O-V-D
0610 NRC RO EXAM 6 RO 295032EA1.01 001lCIAlTlG1IE0I-311295032EA1.01lIRO/SRO/MODIFIED 11117/07 Given the following plant conditions:
* Unit 2 experiences a Main Steam Line break at full powe * Both Inboard and Outboard Main Steam Isolation Valves (MSIVs) on the 'B'
steam line fail to isolate. However, the reactor scrams and ALL rods fully inser * Steam Leak Detection Panel 9-21 indications are as follows:
- 2-TI-1-60A 320 of - 2-TI-1-60B 323 of - 2-TI-1-60C 337 of - 2-TI-1-60D 318 of * NO other temperature indications are alarming at this tim Which ONE of the following describes the appropriate operator actions and the reason for those actions?
REFERENCE PROVIDED Emergency depressurize the reactor due to t~o (2) areas being above Max Safe in EOI-3, "Secondary Containment Control."
B." Enter 2-EOI-1, "RPV Control" and initiate a Reactor Scram due to one (1) area being above Max Safe in EOI-3, "Secondary....Containment Control." Rapidly depressurize the reactor due to one (1) area above Max Safe and one (1) area approaching Max Safe in EOI-3, "Secondary Containment ControL" Enter 2-GOI-100-12A, "Unit Shutdown" and commence a normal shutdown / cooldown due to a primary system discharging outside Primary Containment.
Friday, February 29, 2008 3:01 :07 AM 132
0610 NRC RO EXAM KIA Statement: 295032 High Secondary Containment Area Temperature I 5 EA 1.01 - Ability to operate andlor monitor the following as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE: Area temperature monitoring system KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the required actions which result from high secondary containment temperatures as indicated by Area Temperature Monitoring instrumentatio References: 2-EOI-3 Flowchart, EOIPM Section O-V-E 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 outcom NRC Exam MODIFIED FROM OPL 171.204 #20 REFERENCE PROVIDED: 2-EOI-3 Flowchart Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Which area(s) are above or approaching Max Safe 2. Based on Item #1 above, determine the appropriate action and the basis for that actio B * correct: A* incorrect: This is plausible because ALL four temperatures provided are greater than 315°F as indicated on Table 3. However, ONLY one indicator applies to an EOI-3 area, therefore ONLY one area is above Max Saf C * incorrect: This is plausible because one area is above Max Safe and given conditions indicate an unisolable leak exists; which implies conditions are degradin However, with NO other temerature indications in alarm, anticipating the requirement to Emergency Depressurize is NOT ppropriat D * incorrect: This is plausible because ALL four temperatures provided are greater than 315°F as indicated on Table 3. However, ONLY one indicator applies to an EOI-3 area, therefore ONLY one area is above Max Safe. In addition, this step is ONLY addressed if Emergency Depressurization will NOT reduce the discharge into Secondary Containment. In this case, it would reduce the discharge.
Friday, February 29, 2008 3:01 :07 AM 133
EOI-3, SECONDARY CONTAINMENT CONTROL BASES EOI PROGRAM MANUAL SECTION O-V-E DISCUSSION: ENTRY CONDITIONS: EOI-3 Entry conditions for this procedure are symptomatic of conditions which, if not corrected, could degrade into an emergency. Adverse affects on equipment operability and conditions that directly challenge secondary containment integrity were specifically considered in the selection of these entry conditions. Following is a description of each entry condition: Area temperature above the maximum normal operating value or Table 3 A secondary containment area temperature above the maximum normal operating value of Table 3, Secondary Containment Area Temperature, is an indication that steam from a primary system may be discharging into secondary containment. As temperatures continue to increase, continued operability of equipment needed to carry out EOI actions may be compromised. High area temperatures also present a danger to personnel since access to secondary containment may be required by actions specified by EOl Maximum normal operating temperature is defined to be the highest value of a secondary containment area temperature expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properl Differential pressure at or above <A.38> inches of water High secondary containment differential pressure is indicative of a potential loss of secondary containment structural integrity, and could result in uncontrolled release of radioactivity to the environment.
( Reactor Zone Ventilation exhaust radiation level above <A.39> High Reactor Zone Ventilation exhaust radiation levels may indicate that radioactivity is being released to the environment when the system should have automatically isolate Refuel Zone Ventilation exhaust radiation level above <A.40> High Refuel Zone Ventilation exhaust radiation levels may indicate radioactivity is being released to the environment when the system should have automatically isolate Floor drain sump water level above <A.41> A secondary containment floor drain sump water level above maximum normal operating level is an indication that steam or water may be discharging into secondary containmen Maximum normal operating floor drain sump water level is defined to be the highest value of secondary containment floor drain sump water level expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properl Area water level above <A.42> Secondary containment area water level above maximum normal operating level is an indication that steam or water may be discharging into secondary containmen Maximum normal operating secondary containment area water level is defined to be the highest value of secondary containment area water level expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properl REVISION 1 PAGE 9 OF 73 SECTION O-V-E
EOI-3, SECONDARY CONTAINMENT CONTROL BASES EOI PROGRAM MANUAL SECTION O-V-E I DISCUSSION: SC/T-6 and SCIT -7* I Step SC/T-6 is a before decision step that has the operator evaluate current and future efforts to lower secondary containment area temperatures, in relation to the current value and trend of secondary containment area temperatures, to determine if a reactor scram is necessary. The before decision step requires that this determination and subsequent actions be performed before any secondary containment area temperature reaches its respective maximum safe operating temperature value provided in Table Maximum safe operating temperature is defined to be the highest temperature at which neither: I) equipment necessary for the safe shutdown of the plant will fail, nor 2) personnel access necessary for safe shutdown of the plant will be prevented. The maximum safe operating temperature value for all secondary containment areas is provided in Table 3, Secondary Containment Area Temperatur This step is reached only when additional actions have been required to reverse an increasing secondary containment area temperature trend. If all secondary containment area temperatures can be maintained below their respective maximum safe operating val.ues, the operator returns to Step SCIT-l. If it is determined that all secondary containment area temperatures cannot be maintained below their respective maximum safe operating values, the operator continues at Step SCIT - Step SCIT-7 is an enter and execute concurrently step that requires the operator to enter EOI-I, RPY Control, at Step RC-I, and to perform the actions concurrently with this procedure. Because this step is prioritized with the miniature before decision step symbol relating to SC/T-6, this action should be performed before any secondary containment area temperature reaches its respective maximum safe operating value .
- Initiation of reactor scram (Step RC-l) before any secondary containment area temperature reaches its respective maximum safe operating value may halt the increase in secondary containment area temperature(s), since the RPY is the only significant source of heat, other than a fire, that could cause secondary containment area temperatures to exceed their respective maximum safe operating values .
REVISION 1 PAGE 27 OF 73 SECTION ON-E
TABLE 3 SECONDARY CNTMT AREA TEMP PA!~EL~ P.t..NEL 9-21 1.1A)( !.I.AX POTENTI.... L ARE.b, ALA.RM WtNDOVv TEMP ELEMENT NORM.t.,L SA,FE ISOLA.nON
{UNLESS NOTED) {UNLESS NOTED) VALUEnF VALLlEnF SOURCES RHR SYS I PUMPS X.... -5'>--3E04 74-95A AL....Rl.1ED 'i50 FCV-7A-41.4e RHR $YS HPUMPS XA...s[~E-4* 74-958 ALb,RMEO 210 FC\,-14-41,4e HPC1ROOM XA-55-3F-iO 7~55A Alb,RMED 270 FCV-l~ . 44, S1 CS SY S 1PUMPS Xt..-5'>--3D-10 71-411>. .A,LP.RMED 190 Fell-71-l.. 3. 39 RClCROOM CSSYS UPtJMPS X4.-5'>--3E-29 7/.H:."98 {PANEL~) ALA.RMED 150 NONE XA.-55-30-10 71-418. C. D .A.Lt..RL1ED 200 FCV-l1-1:.3 TOP OF TORUS X......s'>--3F*10 1~5E. ALA.RMED 240 FeIl-I?'-2. 3.81 Xt..Q"':sE-A 74-9[,(;. H A!..ARMED 240 FCV-;4-47,48 A!..ARMED 315 MSIVs STEAl.! TUNNEL {RB) XA.-55-3D-24 1-6M {P.t..NEL 9-3)
FCV-71-2. 3, FCV-OO-1, 2:. 12 DWACCESS X.... ..s'>--3E-4 14-95£ ALA.RMED 170 FCII-74-47.4$ RBU5S5W XA.-55-5S-32 {PANeL 9-5) ~~'A.B; A!..ARMED
\70 FCV.-Q9-1. 2.12 (ROCU PIPE TRENCH) XA,-5C--5S-S3 (PANEL 9-5) tA.UX1NST ROOM) ALt..RMED ROCU H. X. ROOM XA..)'>::~3D-11 OO-29F. G, H ..S,Lb,RMED 220 FCV..Q9-';,2.12 RWCU PUMP A Xb,-6~~D-n OO-.29D ALA.RMED 215 FeV..Q9-1.2. 12 RVv'CU PUMP B X4,..s5-~.D-l1 OO-29E .t..Lt..RLIEO 215 FCV..Q9-1.2.12 RB EL 693 Xf".q.,.:SE-4 7<1*95C. D ALt;RMED 195 FCV-,,!4-41.4e RBEt621 XA-5'>--3E-A ld-9SF ALA.RLtEO 155 FCV-A3-1:3. 1&
.J
...J .J Oil
.J !!t~ i~
'ii' J; .! ~ #l:t ~l ~§ i;
n
,~
i~ Ir: :ii I~ I i I II . .,'"
~ '"
+
WHILE EXECUTING THE FOLLOWING STEPS: IF THEN EMERGENCY RPV DEPRESSURIZATION IS ANTICIPATED RAPIDLY DEPRESSURIZE THE RPV AND WITH THE MAIN TURB BYPASS VL Vs )-- THE REACTOR WILL REMAIN SU8CRlTlCAL WITHOUT BORON UNDER ALL CONDITIONS (SEE NOTE) IRRESPECTIVE OF COOLDOWN RATE L RCIP~3
I I REF ROVIDE TO CAN I ~
et) -ow* r--~' III .iilll II i~! i"
'~!fi' ~ .. Iii; ,iili' ..nilIIi ! 'i l 'I'" '" ,ii' HI !
i ,.1 II'
:i -'I '1'1'[' ,I' '+11 **'*1*
1l ! I; I'
~
I
~! ;1\ !!!III!
I'll II! !
-! ! II!I! I ilill ~ ~ :!!i ~ ~U 'I! '"
iillll! il ~ , I-
:"i!~ IIi'i ~ ,~i ~1 >-'-~-~ II---+---->I Ililllli,!! i!~af~ l ilil~ : iiiI! ~
Illh i I: tile f
~ "1 ~
w! CW)
-*
ow
0610 NRC RO EXAM 64. RO 295033EA2.01 001lCIA/TlG2/SCIR/1295033EA2.01llRO/SROIBANK Given the following plant conditions:
* Unit 2 is at 100% rated powe * A Reactor Water Cleanup (RWCU) drain line cracks and is spilling into the Reactor Buildin * Area Radiation Monitors in the Reactor Building read as follows:
Reactor Building Elevation 593 1100 mR/hr Reactor Building Elevation 565 West 800 mR/hr Reactor Building Elevation 565 East 850 mR/hr Reactor Building Elevation 565 Northeast 1050 mR/hr All other Reactor Building areas NOT ALARMED
* Approximately one (1) minute later, RWCU is successfully isolate Which ONE of the following describes the required action that MUST be directed by the Unit Supervisor and/or Shift Manager?
REFERENCE PROVIDED Enter 2-EOI-1, "RPV Control" and initiate a Reactor Scram due to two (2) areas being above Max Safe in EOI-3, "Secondary Containment Contro!." Scram the reactor and Emergency Depressurize the reactor due to two (2) areas being above Max Safe in EOI-3, "Secondary Containment Contro!." Rapidly depressurize the reactor, to the Main Condenser, with the Main Turbine Bypass Valves due to an Emergency Depressurization being anticipate D..... Enter 2-GOI-1 00-12A, "Unit Shutdown" and commence a normal shutdown / cooldown due to two (2) areas above Max Safe with the source of the leak isolated.
Friday, February 29, 2008 3:01 :07 AM 134
0610 NRC RO EXAM KIA Statement: 295033 High Secondary Containment Area Radiation Levels I 9 EA2.01 - Ability to determine andlor interpret the following as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS: Area radiation levels KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the required actions which result from high secondary containment radiation levels as indicated by Area Radiation Monitoring instrumentatio References: 2-EOI-3 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 outcom NRC Exam REFERENCE PROVIDED: 2-EOI-3 Flowchart Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Which area(s) are above or approaching Max Safe 2. Based on Item #1 above, determine the appropriate actio D - correct: A - incorrect: This is plausible because this action requires at least one area greater than Max Safe. However, this is not appropriate since the source of the leak is isolate B - incorrect: This is plausible because this action requires two areas greater than Max Safe. However, this is not appropriate since the source of the leak is isolate C - incorrect: This is plausible because this action requires at least one area greater than Max Safe and another area approaching Max Safe. However, this is not appropriate since the source of the leak is isolated.
Friday, February 29, 2008 3:01 :07 AM 135
. TABLE 4 SECONDARY CNTMT AREA RADIATION APPLICABLE MM MM POTENTIAL AREA RADIATlON NORMAL SAFE ISOLATION INDICATORS VALLlE. MRfHR VALUE. MR/HR SOURCES RHR SYS I PUMPS 90-Z.l>, 1000 FCV-74-47,4a AlARMED RHR 51'S II PUMPS gO-2M 1000 FCY-74-47.4a AlARMED HPClROOM 90-24A ;(}'O FCY*13-2, ~. 44. S 1 ALARMED CSSYSI PUMPS 90-26A AlARMED 1000 KV-71-2.3.39 RClCROOM CS SYS II PUMPS 00-27A ALARMED 100) NONE FCV-f3-2. 3, 81 TO? OF TORUS 100) 9O-2M AlARMED FCV-74-47, 4a GENERAL AREA FCV-71- R8EL,?6SW OO-ZOA 100) FCV~1.2.12 AlARMED SDVVENTS& DRAINS
~'\..21A S!)VVENTS a DRAINS RBEL&S5E AtJo.RMED 1(}:J)
R8EL!..s5NE 9~23A AlA,:U,IED 100) NONE TIP ROOM 9O-22A A!.ARMED 1(}'.@ TI? BALL VALVE RBELI393 00-1aA.. 14A AlARMED 10" FCV-74-47.48 RB EL 621 90-9) AlARMED 100) FCV-43-13, 14
~'\.4A AtJo.RMED 100:) NONE RECIRC MG SETS REFUELROOR ~1A.2A.3A A!.A'qMED 10:J) NONE
...... * . . . . .
/' /'
l:) WILL EMERGENCY tW L DEPRESSURIZ.ATION REDUCE DISCHARGE INTO SEC ON DlJ'<Y CNTMT SCiR'-4 YES
/ /~_O"/*l../
ARE
/"/RADL~,TK)N l VLs IN 20R:'1~RE~REAS
,"
~.
~
./
NO
..,.
A&O\" E ,,1ft.), SAFE ~
(Tft,BLE 4) ~
/,/
SC/ "'~/~ES L L SCiR-tO
/,"-
\\,;'-1£14 R,'!"DlftTION LVLs IN 2 OR MORE AREt..S AREABOVE hI.AX &A.FE(TABLE.1l}
\- TI1EN CONTINUE sctR~1 EMERGENCY RPV D.EPRE5SUJUZATlON .15 REQUIRED (EOI-1. RClP-4: C1<2:; Ci*21; C~>'l)
L SCiR...e L .... ... .,.
.
WHILE EXECUTING THE FOLLOWING STEPS: IF THEN EMERGENCY RPV DEPRESSURIZATION IS ANTICIPATED AND RAPIDLY DEPRESSURIZE THE RPV
"VITH THE MAIN TURB BYPASS VLVs THE REACTOR WILL REMAIN SUBCRITlCAL IRRESPECTIVE OF COOLDOWN RATE WITHOUT BORON UNDER ALL CONDITIONS (SEE NOTE) . L
E* MINATION REFERENCE PROVIDED TO CANDIDATE
(t)
o-I w
~ " t J
~i~HJ~~f
/
// v~ ~
i
'-I !
I,i"l i t Ige! t, ..J i :' I
-"-l ~ 1 _.I~II)~__
J ,",
;j" i '--\ l~!/--!llj J-<'q!\I/ t i \:)~I '"\J '* lin '" '"1'1I \!y !,- -
v ~- !!. . I U ! CW)
---------------~--~-~.--------------~---
___________L.________~______~_..J -ow*
0610 NRC RO EXAM 65. RO 295035EA2.02 001!C/A/EOI/EOI-3/S85/295035EA2.02//RO/SRO/BANK Given the following plant conditions:
* Unit 2 is at 100% powe * During the backwash of a Reactor Water Cleanup (RWCU) Demineralizer, the Backwash Receiving Tank rupture * The RWCU system has been isolate * Secondary Containment conditions are as follows: - ALL Reactor and Refuel Zone radiation monitors trip on high radiatio ONLY Standby Gas Treatment (SGT) train 'C' can be started. It is operating at 10,000 scfm and taking suction on the Refuel and Reactor Zone Refuel zone pressure: (-)0.12 inches of water - Reactor zone pressure: (+)0.02 inches of water - AREA RADIATION LEVELS RB EL 565 W, 565 E, 565 NE: 250 mr/hr RB EL 593 upscale RB EL 621 upscale Which ONE of the following describes the required action and the type of radioactive release in progress?
REFERENCE PROVIDED A. Initiate a shutdown per 2-GOI-1 00-12A, "Unit Shutdown." An Elevated radiation release is in progres B~ Initiate a shutdown per 2-GOI-1 00-12A, "Unit Shutdown." A Ground-level radiation release is in progres C. Scram the reactor and Emergency Depressurize the RPV. An Elevated radiation release is in progres D. Scram the reactor and Emergency Depressurize the RPV. A Ground-level radiation release is in progress.
Friday, February 29,20083:01 :07 AM 136
0610 NRC RO EXAM KIA Statement: 295035 Secondary Containment High Differential Pressure EA2.02 - Ability to determine andlor interpret the following as they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE: Off-site release rate: Plant-Specific KIA Justification: This question satisfies the KIA statement by requiring the candidate to correctly identify the type of off-site release and required actions due to high differential pressure in the secondary containmen References: 2-EOI-3 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 solve a problem. This requires mentally using this knowledge and its meaning to resolve the proble NRC Exam Friday, February 29, 2008 3:01 :07 AM 137
0610 NRC RO EXAM REFERENCE PROVIDED: 2-EOI-3 flowchart Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Which area(s) are above or approaching Max Saf . Based on Item #1 above, determine the appropriate actio . Whether plant conditions indicate an elevated or ground-level releas NOTE: EOI-3 steps SC/R-8 and SC/R-9 apply, requiring shutdown per 2-GOI-1 00-12A because two (2) or more areas are above max safe rad levels; but, a primary system is NOT discharging to the Reactor Building. Insufficient Reactor Building-to-atmosphere dp (greater than -0.25 inches of water) indicates loss of secondary containment integrity. The positive Reactor Zone pressure is causing an unmonitored and uncontrolled ground-level release of radioactive contaminant B - correct: A - incorrect: The release from the Reactor Building is NOT elevated. This is plausible because the required actions are correct except the differential pressure results in a ground-level releas C - incorrect: Conditions DO NOT warrant a scram at this point. In addition, the release from the Reactor Building is NOT elevated. This is plausible if the candidate fails to recognize that a primary system is NOT discharging to the Reactor Buildin D - incorrect: Conditions DO NOT warrant a scram at this point. This is plausible if the candidate fails to recognize that a primary system is NOT discharging to the Reactor Building.
Friday, February 29, 2008 3:01 :07 AM 138
TABLE 4 . SECONDARY CNTMT AREA RADI.ATION AF>PUCASlE MAX MAX POTENTiAl AREA RADIATION NORMAL SAFE ISOLATION INDiCATORS VALUEMR/HR VALUEMR/HR SOURCES RHR 51'S I PUMPS 90-2GA 1000 FeY*74..!7.4. ALARMED RHR SVS II PUMPS 90-28A 100Q FCY-74-4 7. 4!.l A!..PJtt.IED HPCIROOM 90-24A 1000 FCV-73-2, 3,44. 81 AlARMED CSSYSl PUMPS 90-26A AlARMED 101,)() Fc\!*71*2. 3,39 RCiCROOM CS SYS 11 PUMPS 9ti*ZlA ALARMED 1000 NONE FCV-73-2. 3, 81 TOP OF TORUS 9!J..29A ALAAMED 1000 FCV-74-47. 4!.l GENERAL AREA FCW71-2.3 RBElOO5W 9!J..20A FCV~1.2.12 A!..P.RMED 1000 SDVVENTS & DRA1NS RBEl$5E 9!J..21A 1000 SDVVsrrS 8. DRAINS ALARMED RBELC+SSNE 9!J..2M AlARMED 1000 f.jONE TIP ROOM 9!J..22A AlARMED 100.000 TlPBALL VALVE RBEl593 9!J..i3A.14A A!..P.RMED 1000 FCV-f4..!7.4!.l RBEl621 90-9A AlAR~;!EO 1000 FCV-43-13. 14 RECIRC MG SETS 00-4A ALARMED 1000 NONE REFUEl flOOR 9!J..1A, 2A, M ALARMED 1000 NONE
----
ISOLATe All SYSTEMS THAT "oRE DISCHARGING INTD THE .~ EXCEPT SYSTEMS REQUIRED TO:
* BE OP£R.... TED BY £Ola QR .. SUPPRESS A FIRE L
SC!R~
/ (0 /
WIll EMERGENCY NO L DEPRESSLIRIZ....TIOO REDUCE mSCHARGE INTO SECONDARY CNTM/' SClR-4 YES L ARE NO L AAOIATtoN LVLs IN lOR MORE AREAS AOOVEMAXSAFE .,,/' tTASLE4) .,,/ SCfR-~ SCf~ L
~,
..!.RE NO L RA,D1ATION L VLa IN 20R MORE AREJoS ABOVE MAX SAFE (fABLE4)
SCf.R-9 YES L L L V"HEN AAOIATION lVt.s IN 2: OR MORE AREAS
-- ARE ABOVE MA.X SAFE (fABLE 4)
THEN CONT1NUE SCJR-7 EMERGENeyRPV DEPRE.55URI2'ATlON 15 REQUIRED
{EOI-i. RCJP-4; C1-2: C1-21; (;5-1)
SCiR-.S
EXAMINATION REFERENCE PROVIDED TO . CANDIDATE
(I')
-owI (I')
---- __ ~_ _ ________..L.____. _____________ _ _L___ _ -
ow I
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. Tech Spec applicability for the listed systems with the given plant conditio NOTE: The distractors are all plausibe since only one system or function is incorrect in each distracto C - correct: A - incorrect: The RBM is NOT required until >27% rated powe B - incorrect: The APRM Hi (120%) is NOT required until Mode D - incorrect: The OPRMs are NOT required until Mode 1 and >25% rated power.
Friday, February 29, 2008 3:01 :07 AM 140
Control Rod Block Instrumentation 3.3. Table 3.3.2.1-1 (page 1 of 1) Control Rod Block Instrumentation APPLICABLE MODES OR FUNCTION OTHER REQUIRED SURVEILLANCE ALLOWABLE SPECIFIED CHANNELS REQUIREMENTS VALUE CONDITIONS 1. Rod Block Monitor Low Power Range - Upscale (a) 2 SR 3.3.2. (e) SR 3.3.2. SR 3.3.2. b. Intermediate Power Range - Upscale (b) 2 SR 3.3.2. (e) SR 3.3.2. SR 3.3.2. High Power Range - Upscale (I),(g) 2 SR 3.3.2. (e) SR 3.3.2. SR 3.3.2. d. Inop (g),(h) 2 SR 3.3.2. NA Downscale (g),(h) 2 SR 3.3.2. (i) SR 3.3.2. . Rod Worth Minimizer 1(c),2(c) SR 3.3.2. NA SR 3.3.2. SR 3.3.2. SR 3.3.2. . Reactor Mode Switch - Shutdown Position (d) 2 SR 3.3.2. NA (a) THERMAL POWER ~ 27% and ~ 62% RTP and MCPR less than the value specified in the COLR.
(b) THERMAL POWER> 62% and ~ 82% RTP and MCPR less than the value specified in the COLR.
(c) With THERMAL POWER ~ 10% RTP.
(d) Reactor mode switch in the shutdown position.
(e) Less than or equal to the Allowable Value specified in the COLR.
(I) THERMAL POWER> 82% and < 90% RTP and MCPR less than the value specified in the COLR.
(g) THERMAL POWER ~ 90% RTP and MCPR less than the value specified in the COLR.
(h) THERMAL POWER ~ 27% and < 90% RTP and MCPR less than the value specified in the COLR.
(i) Greater than or equal to the Allowable Value specified in the COLR.
BFN-UNIT 1 3.3-20 Amendment No. ~262 September 27, 2006
RPS Instrumentation 3.3.1.1 ') Table 3.3.1.1-1 (page 1 of3) 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 . Intermediate Range Monitors Neutron Flux - High 2 3 G SR 3.3.1. " 120/125 SR 3.3.1. divisions of full SR 3.3.1. scale SR 3.3.1. SR 3.3.1. SR 3.3.1.1.14 5(a) 3 H SR 3.3.1. " 120/125 SR 3.3.1. divisions of full SR 3.3.1. scale SR 3.3.1.1.14 Inop 2 3 G SR 3.3.1. NA SR 3.3.1.1.14 5(a) 3 H SR 3.3.1. NA SR 3.3.1.1.14 Average Power Range Monitors Neutron Flux" - High, 2 3(b) G SR 3.3.1. ,,15% RTP Setdown SR 3.3.1. SR 3.3.1. SR 3.3.1.1.13 SR 3.3.1.1.16 Flow Biased Simulated 3(b) F SR 3.3.1. " 0.66 W Thermal Power - High SR 3.3.1. + 66% RTP SR 3.3.1. and" 120% SR 3.3.1.1.13 RTP(c) SR 3.3.1.1.16 Neutron Flux - High 3(b) F SR 3.3.1. ,,120% RTP SR 3.3.1. SR 3.3.1. SR 3.3.1.1.13 SR 3.3.1.1.16 (continued)
(a) With any control rod withdrawn from a core cell containing one or more fuel assemblie (b) Each APRM channel provides inputs to both trip system (c) [0.66 W + 66% - 0.66 /::" VV] RTP when reset for single loop operation per LCO 3.4.1, "Recirculation Loops Operating."
BFN-UNIT 1 3.3-6 Amendment No. 236, 262, 269 March 06, 2007
RPS Instrumentation 3.3. Table 3.3.1.1-1 (page 2 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 . Average Power Range Monitors (continued) Inop 1,2 3(b) G SR 3.3.1.1.16 NA ut-Of-4 Voter 1,2 2 G SR 3.3.1. NA SR 3.3.1.1.14 SR 3.3.1.1.16 OPRM Upscale 3(b) SR 3.3.1. NA SR 3.3.1. SR 3.3.1.1.13 SR 3.3.1.1.16 SR 3.3.1.1.17 Reactor Vessel Steam Dome 1,2 2 G SR 3.3.1. " 1090 psig Pressure - High(d) SR 3.3.1. SR 3.3.1.1.10 SR 3.3.1.1.14 Reactor Vessel Water Level - 1,2 2 G SR 3.3.1. ;:: 528 inches Low, Level 3(d) SR 3.3.1. above vessel SR 3.3.1.1.13 zero SR 3.3.1.1.14 Main Steam Isolation Valve - 8 F SR 3.3.1. ,,10% closed Closure SR 3.3.1.1.13 SR 3.3.1.1.14 Drywell Pressure - High 1,2 2 G SR 3.3.1. ,,2.5 psig SR 3.3.1.1.13 SR 3.3.1.1.14 Scram Discharge Volume Water Level - High Resistance Temperature 1,2 2 G SR 3.3.1. ,,50 gallons Detector SR 3.3.1.1.13 SR 3.3.1.1.14 5(a) 2 H SR 3.3.1. ,,50 gallons SR 3.3.1.1.13 SR 3.3.1.1.14 (continued) (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.
(b) Each APRM channel provides inputs to both trip systems.
(d) During instrument calibrations, if the As Found channel setpoint is conservative with respect to the Allowable Value but outside its acceptable As Found band as defined by its associated Surveillance Requirement procedure, then there shall be an initial determination to ensure confidence that the channel can perform as required before returning the channel to service in accordance with the Surveillance. If the As Found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperabl Prior to returning a channel to service, the instrument channel setpoint shall be calibrated to a value that is within the acceptable As Left tolerance of the setpoint; otherwise, the channel shall be declared inoperabl The nominal Trip Setpoint shall be specified on design output documentation which is incorporated by reference in the Updated Final Safety Analysis Report. The methodology used to determine the nominal Trip Setpoint, the predefined As Found Tolerance, and the As Left Tolerance band, and a listing of the setpoint design output documentation shall be specified in Chapter 7 of the Updated Final Safety Analysis Report.
BFN-UNIT 1 3.3-7 Amendment No. 269 234, 262, 259, 257,258, 266 March 06, 2007
0610 NRC RO EXAM 67. RO GENERIC 2.1.16 001IMEM/T31IB17/G2.1.16//RO/SROINEW 10116107 Which ONE of the following announcements is an INAPPROPRIATE use of the Plant Paging System in accordance with OPDP-1, "Conduct of Operations?" "Shift Manager dial 2391. Shift Manager dial 2391." "Operations will be starting the 2 Alpha RHR pump."
C." "All personnel evacuate the Unit 2 Reactor Building due to high radiation. This is a drilL" "There is a fire in the Unit 2 Shutdown Board Room. I repeat. There is a fire in the Unit 2 Shutdown Board Room."
KIA Statement: Conduct of Operations 2.1.16 Ability to operate plant phone, paging system, and two-way radio KIA Justification: This question satisfies the KIA statement by requiring the candidate to demonstrate specific knowledge of the use of the Plant Paging System while communicating with plant personne References: OPDP-1 Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam Friday, February 29, 20084:44:41 AM 141
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the following: 1. The requirements associated with making Page Announcements per OPDP- . Whether the announcement meets those requirements C - correct: The line "This is a drill" is required at the beginning AND end of each communication during drills / exercises. In addition, an announcement of such urgency should be repeate A - incorrect: This is plausible since the page is repeated. However, there may NOT be a requirement to repeat the announcement, but it is NOT an inappropriate actio incorrect: This is plausible since the page is NOT repeated. However, repeating pages for a normal operation is NOT require D - incorrect: This is plausible because it is an expected announcement during a fire.
Friday, February 29,20084:44:41 AM 142
TVAN Standard Conduct of Operations OPDP-1 Department Rev.OOOa Procedure EC3_ge ?§ of 1JlL Appendix I (Page 3 of 5) Communications Use equipment noun names and/or identification (ID) numbers to describe a componen The use of sign language is undesired but maybe used when verbal communications is not practica Take time when reporting abnormal conditions. Speak deliberately, distinctly and calmly. Identify yourself and watch station or your location. Describe the nature and severity of the problem. State the location of the problem if appropriate. Keep the communication line open if possible or until directed otherwis The completion of directed actions should be reported to the governing station, normally the control roo Require other plant personnel (including contractors) conducting operational communication to do so in accordance with this procedur If there is any doubt concerning any portion of the communication or task assigned, resolve it before taking any actio When making announcements for drills or exercises begin and end the announcement with "This is a Drill." Emergency Communications Systems When personnel are working in areas where the public address (PA) system or emergency signals cannot be heard, alternate methods for alerting these persons should be devise Flashing lights, personal pagers that vibrate and can be felt, and persons dedicated to notifications are examples of alternate method . PA System Use of the plant PA system shall be limited to ensure it retains its effectiveness in contacting plant personnel. Excessive use of the PA system should be avoided. Plant telephones and other point-to-point communications channels should be used in lieu of the PA system whenever practica The plant PA system may be used in abnormal or emergency conditions, to announce change of plant status, or give notification of major plant events either in progress or anticipate TVAN Standard Conduct of Operations OPDP-1 Department Rev. 0008 Procedure Page 56 of 103 Appendix I (Page 4 of 5) Communications When using the plant PAsystem:
(1) Speak slowly and deliberately in a normal tone of voic (2) When announcements of abnormal or emergency conditions are made, they shall be made at least twic (3) When making announcements for drills or exercises begin and end the announcement with "This is a Drill." . Plant Telephones When using Plant telephones: Identify yourself and watch statio When trying to make contact with the main Control Room, if the message is of a routine nature, the sender should hang up when the main Control Room fails to answer after the fifth ring to avoid unnecessary Control Room noise. The phone shall be allowed to ring until answered if the information is important to Operation During times when the DO NOT DISTRUB (DND) function has been used by MCR personnel, follow the directions on the recording as appropriat When making announcements for drills or exercises begin and end the announcement with "This is a Drill." Radio/phone Communication Radio/phone usage shall not be allowed in areas where electronic interference with plant equipment may resul When making announcements for drills or exercises, begin and end the announcement with "This is a Drill." Sender should identify themselves by watch statio Three way communications should be use Clear concise language should be used since radio/phone contact does not have the advantage of face to face communicatio NRC RO EXAM 68. RO GENERIC 2.1.18 001/MEM/T3/12.1I/GENERIC 2.1.181IRO/SROINEW 10/16/07 Which ONE of the following is an INEFFECTIVE use of the phonetic alphabet in accordance with OPDP-1, "Conduct of Operations?" "Start 2 Alpha RHR pump per 3-01-74." "Place Golf IRM in Bypass per 1-01-92-Bravo." "Transfer 2 Alpha 480 Volt Shutdown Board to Alternate."
D." "Place Romeo Papa Sierra 2 Alpha on Alternate per 2-0scar-lndia-99."
KIA Statement: Conduct of Operations 2.1.18 Ability to make accurate, clear and concise logs, records, status boards, and report KIA Justification: This question satisfies the KIA statement by requiring the candidate to demonstrate knowledge of the requirements related to verbal communications or reports during shift operation References: OPDP-1 Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam Friday, February 29, 2008 3:01 :07 AM 143
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the requirements for use of the phonetic alphabet and apply that knowledge to the given communication NOTE: Each distractor is plausible because they all contain at least one use of the phonetic alphabe D - correct: The use of the phonetic alphabet for common acronyms, such as RPS, is not required and could reduce the effectiveness of the communicatio A - incorrect: This communication is appropriat B - incorrect: This communication is appropriat C - incorrect: This communication is appropriate.
Friday, February 29, 2008 3:01 :07 AM 144
TVAN Standard Conduct of Operations OPDP-1 Department Rev.OOOS ) Procedure P~e 54 of 103 Appendix I (Page 2 of 5) Communications The receiver repeats back the message to the sender. The repeat back can be verbatim or functional. In many cases a functional repeat back best communicates the receivers understanding of the message. This can be done in several ways to accomplish the desired goals. For example the sender might say, "Bob, report RCS pressure and trend." The receiver could respond in either of two way (1) The receiver could respond with, "Report RCS pressure and trend. RCS pressure is 2250 psig and stable."
Or (2) The receiver could respond with, "RCS pressure is 2250 psig and stable." The sender verbally acknowledges that the receiver correctly understood the message. The verbal acknowledgement can be simple such as, "That is correct". If the sender has requested and received information then the sender shall provide either verbatim or functional repeat back to demonstrate his understanding of the receiver's message. For the example above the sender could respond with, "I understand 2250 and stable." Phonetic Alphabet The phonetic alphabet is a tool to improve communications. In general, operations communication should use the phonetic alphabet except when well established acronyms describe the subject. If use of phonetic alphabet will reduce effectiveness of communications then it should not be used. The following are examples of when the phonetic alphabet should not be used: It is not desirable to use Romeo-Charlie-Sierra to describe the RCS (Reactor Coolant System). If a procedural step is written using acronyms, it may be read and ordered as suc If a component tag or label is written using acronyms then the acronyms may be use . General Standards All communications shall be clear, concise, and precise. All operational communications shall be conducted in a formal and professional manner. In all communications, the sender and intended receiver should be readily identifiabl NRC RO EXAM 6 RO GENERIC 2.2.13 001/MEMlT3/l0.217118/GENERIC 2.2. 13/3.6/3.8/RO/SROINEW 10/16/07 Which ONE of the following describes the requirements when placing a clearance on Air-Operated Valves (AOVs)? An Air-Operated Valve that fails _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ "OPEN" on loss of air SHALL NOT be used for blocking purpose B." "OPEN" on loss of air, SHALL be held closed with a gagging device that is tagged as a clearance boundar "AS-IS" on loss of air SHALL NOT be used for blocking purposes until it is verified closed and a gagging device installe "CLOSED" on loss of air SHALL NOT be considered closed for blocking purposes unless it is held closed with a gagging devic KIA Statement: Equipment Control 2.2.13 Knowledge of tagging and clearance procedures KIA Justification: This question satisfies the KIA statement by requiring the candidate to demonstrate knowledge of the Clearance and Tagging requirement References: SPP 1 Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam Friday, February 29, 2008 3:01 :07 AM 145
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly, the candidate must determine the requirements for Clearance and Tagging procedure, SPP-10.2 and apply that knowledge to the given condition B - correct: A - incorrect: This is plausible since using a "Fail-Open" valve presents a difficult problem. However, SPP-10.2 provides specific additional guidance to support the allowance of their use as a clearance boundar C - incorrect: This is plausible because, in most cases, it is true. However, SPP-1 provides specific guidance and controls to allow using them as a clearance boundary under a specific condition that the clearance be considered "working on energized equipment".
D - incorrect: This is plausible since a locking device would ensure the valve does NOT open. However, SPP-10.2 requires the air supply that actuates the valve be mechanically or electrically isolated.
Friday, February 29, 2008 3:01 :07 AM 146
NPG Standard Clearance Procedure to SPP-1 Programs and Safely Control Energy Rev. 0010 j Processes Page 50 of 66 Appendix E (Page 1 of 2) Special Requirements for Mechanical Clearances REQUIREMENTS An air-operated valve that fails closed must have its air supply electrically or mechanically isolated, depressurized, and the valve visually checked-to-be-closed by local or remote indication. The air supply energy-isolating devices must be tagge An air-operated valve that fails "as is" shall be closed and mechanically restrained. Its air supply should be electrically or mechanically isolated, depressurized, and the valve visually checked to be closed by local or remote indication. The air supply energy-isolating devices and mechanical restraint must be tagge In cases where it is not possible to physically secure an air operated valve that fails
"as-is" in the closed position, the valve will be tagged closed by applying closing air to the valve diaphragm by the use of the solenoid valve air overrides and tagging both the hand-switch in the closed position and the solenoid valve air overrides. Prior to allowing work to begin, the equipment will be drained and de-pressurized to ensure the boundary valves are holding. This condition will be noted in the remarks section of the clearance sheet to inform PAE/Authorized Employee(s) that pressurized air is required to ensure the valve remains closed. This work is considered "working on energized equipment" and must be approved by the management official in charg Pressure controlled valves, relief valves, and check valves will not be used as isolation boundary valves under normal conditions. Where such a valve does not have an external means of physical restraint, the work is considered "working on energized equipment" and must be approved by the management official in charg The following instructions govern the use of freeze plugs The clearance should be in place, but not issued, before establishing the freeze plu . The need for the freeze plug should be identified onthe Remarks Section of the clearance sheet. The freeze plug should not be listed as a device held on the clearance sheet. The establishment and maintenance of the freeze plug shall be in accordance with approved procedures or work document . The freeze plug must be attended by qualified personnel to ensure that it is maintained intact until all work is complete and the proper Post Maintenance Tests (PMTs) are performe . If the clearance must be released to allow performance of a PMT, the equipment must be retagged before allowing the freeze plug to thaw. This will prevent migration of a portion of the plu NRC RO EXAM 70. RO GENERIC 2.2.33 OOl/C/A/SYS/RWMIIG2.2.33/RO 2.5/RO/SRO/BANK Given the following plant conditions: - A reactor startup is in progress with power currently at 3% - Rod Worth Minimizer (RWM)is latched into Group 8 (12 control rods) - Group 9 rods are the same rods as Group Sequence Control: ON - Group 8 Limits: 08-12 - Group 9 Limits: 12-16 Which ONE of following describes when the RWM will automatically latch up to Group 9? ALL rods EXCEPT 3 in group 8 are withdrawn to the group Withdraw Limit, and a rod in group 9 is selecte B." ALL rods in group 8 have been withdrawn to the group 8 Withdraw Limit and a rod in group 9 has been selecte ALL rods EXCEPT 1 in group 8 are withdrawn to the group Withdraw Limit and a rod in group 9 has to be selected and move The last rod in group 8 is withdrawn to the group 8 Withdraw Limit and the in-sequence rod in group 9 has NOT been selecte KIA Statement:
Equipment Control 2.2.33 Knowledge of control rod programmin KIA Justification: This question satisfies the KIA statement by requiring the candidate to recognize and apply limitations on control rod programming enforced by the Rod Worth Minimizer progra References: Lesson Plan OPL 171.024 Rev. 13 pages 13 - 15 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 solve a problem. This requires mentally using this knowledge and its meaning to resolve the proble NRC Exam Friday, February 29, 2008 3:01 :07 AM 147
OPL 171.024 Revision 13 Page 14 of 53 INSTRUCTOR NOTES (b) After returning the RWM to service:
(1) Group 7 will be the latched group (2) Rod 30-03 will be displayed as a withdraw erro (3) The withdraw block status indicators will indicate a withdrawal block condition on the RWM system displays and RWM switch pane (4) No other control rod may be inserted or withdrawn until the withdraw error rod from Group 8 (30-03)
is corrected. It can only be inserte (c) The proper way to correct the NOTE: Upon select out of sequence condition is to of rod 30-03, an insert the withdraw error rod (30- RWM system 03) to position "04". message will be generated indicating This removes the withdraw error; a target position of leaves group 7 as the latched notch "04" for this group, and removes the control rod withdraw block indications on the RWM system displays and RWM switch panel.
11. Automatic Latching Up/Down Obj. V.B.10 The automatic latching process depends on whether or not RWM Sequence Control is ON or OFF. Sequence Control is normally selected (ON) and enforces a specific order to pull rods within a latched grou When operating below the LPSP with NOTE: Latching sequence controIION", latching to the next within Transistion higher or next lower rod group is done Zone will be internally by the RWM program only after a discussed late rod in the next group is selecte NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: B - correct: A - incorrect: This is plausible because the RWM normally.~~'1;._~~.Q('§,
~1t~"l"""{,~J$,j'.~~~'~ however it will NOT latch up to a higher group LTh-cleyil:\
tnls'conaitJon5ecause'~'~""Mree rods are more than one notch from the withdraw limi C - incorrect: The selected control rod does NOT have to be moved to latch to Group 9. This is plausible because the RWM will latch to the highest group with one rod past the insert limit if the RWM is latching to a group from an unknown condition. Since this is a known condition, the RWM will latch to Group 9 without moving the selected ro D - incorrect: The RWM will NOT latch to the next group until the correct rod is selected in Group 9; because Sequence Control is ON. This is plausible if Sequence Control is OFF. Under that condition, the RWM only looks for rods within the Group and NOT within a specific sequence. With Sequence Control in OFF, the RWM will latch up to Group 9 as soon as the last rod reaches the withdraw limit.
Friday, February 29, 2008 3:01 :07 AM 148
OPL 171.024 Revision 13 Page 13 of 53 INSTRUCTOR NOTES (8) Upon demand by the operator via the Scan/Latch request functio (9) Following correction of Insert or Withdraw Errors.
d. The latched group is the highest group which can be achieved without producing an active insert block conditio (1) The RWM system will latch to the highest sequence with:
(a) (b) (2) Example: Relatch at an intermediate power level (a) Assume that RWM has been out of service and rods have been moved out of sequenc The following rod distribution exists: (1 ) All rods in Group 1 thru These 3 rods would 7 are at their withdraw cause an insert limit, except rods 30- block if GP 8 were 35, 38-43 and 38-27 latche (GP. 7) which are at position 0 (2) All rods in Groups 8 and above are at their insert limit (04) except for rod 30-03 (GP 8)
which is at position 0 (3) No rod is selected
OPL 171.024 Revision 13 Page 15 of 53 INSTRUCTOR NOTES (1 ) The program will latch down (latch the NOTE: Will latch next lower group) when all the rods in down if insert errors the presently latched group have been in GP is lower than inserted to the group insert limit and a latch G rod in the next group is selecte (2) The program will latch up (latch the Will latch up next higher group) upon selection of a provided that the rod within the next higher group number of insert provided that only 2 insert errors or less errors produced will result from within the current latched not give an insert group and/or any lower group block.
c. When sequence control is NOT selected, (OFF), latching automatically occurs based on rod movement within repeating BPWS banked groups (ex: 2/3/4/5/6 and 7/8/9/10/11/12).
(1) For example, if the rods in a group (G ) are the same rods as in the next higher group (GP. 5), the RWM will NOT latch up based solely upon control rod selection. Latch up to Group 5 will automatically occur when any of the rods Group 4 are moved to a position defined for Group 5 provided that <3 insert errors would resul (2) If the rods in a group (GP. 5) are the With rods at both a same rods as the next lower group GP 4 and GP 5 (GP. 4), the RWM will not latch down defined position, the based solely upon control rod selectio latched GP after a Latch down to the next lower RWM movement will be group will generally occur in this case the GP moved int based upon movement of any of the rods within the group to a position defined for the next lower RWM grou (3) If the next rod group is NOT repeating, then latching occurs when the next rod is selecte NRC RO EXAM 71. RO GENERIC 2.3.10 001lCIA/T3/GENERICICIA/G2.3.101IRO/SROIBANK Given the following conditions at a work sit Airborne activity: 3 DAC Radiation level: 40 mrlhr Radiation level with shielding: 10 mrlhr Time to place shielding: 15 minutes Time to conduct task with respirator: 1 hour Time to conduct task without respirator: 30 minutes Assume the following:
- the airborne dose with a respirator will be zero (0).
- a dose rate of 40 mrlhr will be received while placing the shieldin all tasks will be performed by one worke shielding can be placed in 15 minutes with or without a respirato Which ONE of the following would result in the lowest whole body dose? Conduct the task WITHOUT a respirator or shieldin Conduct the task with a respirator and WITHOUT shieldin Place the shielding while wearing a respirator and conduct the task with a respirato D.oI Place the shielding while wearing a respirator and conduct the task WITHOUT a respirato KIA Statement: Radiation Control 2.3.10 Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposur KIA Justification: This question satisfies the KIA statement by requiring the candidate to calculate the expected exposure for a job and determine the correct precautions and radiological controls required to minimize exposur 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 outcom NRC Exam Friday, February 29,20083:01 :07 AM 149
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: This question requires the candidate to calculate the exposure received for each of the four options in the distractors. Plausibility of distractors is based upon either calculation errors or misapplication of ALARA principles; and is based entirely on the type of decision which must be made while performing duties as a Licensed Operator. Using the calculation below, the candidate must correctly perform the analysis and apply ALARA principles to select the correct answe Calculations required: 3 DAC x 2.5 mr/DAC X 0.5 hours =3.75 mr a. 10 mr placing shielding, 10 mr conducting task, zero airborne =20 mr b. 10 mr placing shielding, 5 mr conducting task, 3.75 mr airborne = 18.75 mr (lowest
=
dose Correct) c. 40 mr conducting task, zero airborne =40 mr d. 20 mr conducting task, 3.75 mr airborne =23.75 mr D - correct: A - incorrect: Would NOT result in dose being maintained ALARA, based upon the calculations abov B - incorrect: Would NOT result in dose being maintained ALARA, based upon the calculations abov C - incorrect: Would NOT result in dose being maintained ALARA, based upon the calculations above.
Friday, February 29,20083:01 :07 AM 150
0610 NRC RO EXAM 72. RO GENERIC 2.3.9 001IMEMlT3/PR.CMPTRlIRO GENERIC 2.3.91IRO/SRO/BANK 11127107 RMS Unit 2 reactor shutdown is in progress and primary containment de-inerting has been authorize Which ONE of the following i " ; st for preventing the simultaneous opening of BOTH the SUPPR CHBR ATM SPLY INBD ISOLATION VLV (2-FCV-64-19) and the DRYWELL ATM SUPPLY INBD ISOLATION VLV (2-FCV-64-18 ) during the performance of this evolution? To prevent the high flow rate from damaging the non-hardened ventilation duct B." To prevent the possibility of overpressurizing the primary containment during a LOC To prevent release of the drywell atmosphere through an unmonitored ventilation flow pat To prevent creating a high differential pressure between the primary containment and the Reactor Buildin KIA Statement: Radiation Control 2.3.9 Knowledge of the process for performing a containment purg KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions and times to correctly determine the process for performing a containment purg References: 2-01-64, Rev.106, section Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam Friday, February 29, 2008 3:01 :07 AM 151
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine the requirements for de-inerting the Primary Containment and their base B - correct: A - incorrect: This is plausible becaue high flowrates would result from both valves being open. However, the vent ducts are designed to accomodate such flowrate C - incorrect: This is plausible since the vent path is unmonitored. However, having both valves open simultaneously provides NO additional path for a releas D - incorrect: This is plausible because the de-inerting lineup raises the differential pressure between the Orywell and Reactor Building. However, the rise is relatively insignificant and well within the design limits.
Friday, February 29, 2008 3:01 :07 AM 152
BFN Primary Containment System 2-01-64 Unit2 Rev. 0106 Page 40 of 194 INFREQUENT OPERATIONS Purging the Drywell and Suppression Chamber with Primary ContainmentPurge Filter Fan NOTES 1) TOE 970823 identified a potential for a bypass flow path to exist between the Drywell and Suppression Chamber when purging the Drywell and Suppression Chamber at the same time (both Therefore, nmary Containmenfpu'rg[nglsreqwretTwifhihe Reacfor'NOf In Cold Shutdown (Mode 4 or 5), the Suppression Chamber and the Drywell are purged separatel ) This section is used when purging both the Drywell and Suppression Chamber concurrently with the Reactor in Cold Shutdown (Mode 4 or 5).
3) When the Reactor is NOT in Cold Shutdown (Mode 4 or 5), the Suppression Chamber and the Drywell are purged separatel [1) REVIEW all Precautions and Limitations in Section 3.0: 0
[2) VERIFY all Prestartup/Standby Readiness requirements in Section 4.0 are satisfie [3) VERIFY the following initial conditions are satisfied: * Drywell vented to less than 0.25 psi REFER TO Section * H2 0 2 analyzers are in service REFER TO 2-01-76 0 * Suppression Chamber vented to less than 0.25 psi REFER TO Section * Reactor Zone Fans in operation with Reactor Zone Supply and Exhaust Fan in fast speed. REFERTO 2-01-30 [4) REQUEST Chemistry to obtain a Drywell sample. REFER TO 2-SI-4.8.B.2- [5) IF sample is within limits of 2-SI-4.8.B.2-6, THEN NOTIFY Shift Manage NRC RO EXAM 73. RO GENERIC 2.4.47 OOllC/A/T3/C4/6/G2.4.471IRO/SRO/BANK Given the following plant conditions: * Reactor pressure is being maintained at 50 psi * Temperature near the water level instrument run in the Drywell is 220 o * The Shutdown Vessel Flooding Range Instrument (LI-3-55) is reading (+)35".
Which ONE of the following describes the highest Drywell Run Temperature at which the LI-3-55 reading (+)35 inches is considered valid? REFERENCE PROVIDED of B..... 250 of of of KIA Statement: Emergency Procedures IPlan 2.4.47 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference materia KIA Justification: This question satisfies the KIA statement by requiring the candidate to use specific plant conditions to determine the correct reactor water level under emergency condition References: 2-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 outcom NRC Exam Friday, February 29,20083:01 :07 AM 153
0610 NRC RO EXAM REFERENCE PROVIDED: 2-EOI-1 flowchart Plausibility Analysis: In order to answer this question correctly, the candidate must use EOI Caution #1 to determine operable RPV water level instrument B - correct: A - incorrect: This is plausible since 200°F is a valid indication, however the question calls for the HIGHEST temperatur C - incorrect: This is plausible if the candidate interpolates the Caution #1 table, however this is NOT permissibl D - incorrect: This is plausible if the candidate interpolates the Caution #1 table, however this is NOT permissible.
Friday, February 29,20083:01 :07 AM 154
CAUTIONS CAUTION #1
- AN RPV WATER LVllNSTRUMENT MAY BE USED TO DETERMINE OR TREND lVL ONLY 'l;VHEN IT READS ABOVE THE MINIMUM INDICATED LV.LASSOC1ATEO WITH THE HIGHEST MAX OW OR SCRUNTEMP .
- IF DW TEA4RS. OR SC AREA TEMPS (TABLE 6), AS APPLICABLE, ARE OUTSmE THE SAFE REGION OF CURVE 8, THE ASSOCIATED INSTRUMENT W\.Y BE UNRELIABLE DUE TO OOIUNG IN THE RU MINIMUM MAX DW RUN TEMP MAXSC INSTRUMENT I RANGE INDICATED (FROM XR-64-50 RUN TEMP LVL OR Tt-64-52AB) (FROM TABLE 6 BELOW* 100
~
151 TO 2(J{! LI-3-5M. B I EMERGENCY -1M) N/A 2t)1 TO 200-HiS TO +&l-130 N/A 251 T030{)
-120 N/A 301 TO 300 U-3-5J ON SCALE N/A BELOW 100 If..3..00 +5 NlA 151 TO 200 LI*3-2{J{) NORMAL
+15 N/A (ITO +00 U-3-253 +2Q NIA U*S-ZQEA, B, C. 0 +30 N/A U<,-52 U-3..fi2A POST ACCIDENT 268 TO +32 I ON SCALE N/A N/A
+10 BELOW 100 N/A
+15 10{) TO 150 N/A SHUTDOWN! +20 151 TO 2(J{! N/A Ll-3-5S ROOOUPl +~ . .2{M'f025Q* . ,: MIA OTO +400 I +40 251 TO 300 N/A
+00 301 TO 300 NtA
+65 351 TO 400 N/A
a~YOIONY~ O~ OaOIAOHd a~NaHa.:laH NOI~YNIII\IYXa .
m ~ - I 0 I \.
* m
.I ...
*
I I I NI m o
...-*
0610 NRC RO EXAM 7 RO GENERIC 2.4.15 001lMEM/T3111GENERIC 2.4. 1511RO/SROINEW 10/16107 Given the following plant conditions:
* Unit 2 has scrammed and multiple control rods failed to inser * The Unit Supervisor has entered EOI-1, "RPV Control", and C-5, "Level/Power ControL" * You have been designated to assist the crew by performing EOI Appendicies as they are assigne Which ONE of the following precludes the use of a handheld radio to communicate with Control Room personnel?
A. EOI Appendix 2 in the 2A Electrical Board Roo B. EOI Appendix 16H at the '2C' 250V RMOV Boar C~ EOI Appendix 1C in the Auxiliary Instrument Roo D. EOI Appendix 1B on the Reactor Building 565/ Elevatio KIA Statement: Emergency Procedures /Plan 2.4.15 Knowledge of communications procedures associated with EOP implementatio KIA Justification: This question satisfies the K/A statement by requiring the candidate to demonstrate knowledge of communication requirements that apply during execution of Emergency Operating Instruction References: OPDP-1 Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam Friday, February 29, 2008 3:01 :07 AM 155
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly the candidate must determine which of the given locations violates the requirements of OPDP-1, Conduct of Operatio C - correct: A - incorrect: This is plausible bacuse of the safety related equiment powered from 2A Electric Board Room; however, radio communication is authorize B - incorrect: This is plausible because of the safety related equipment fed from '2C' 250V RMOV Board; however, radio communication is authorize D - incorrect: This is plausible because of the proximity of the RPV level instrumentation; however, radio communication is authorized.
Friday, February 29, 2008 3:01 :07 AM 156
EOI PROGRAM MANUAL USER'S GUIDE FOR EMERGENCY OPERATING INSTRUCTIONS SECTION O-VIII-A _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _~_ _ __ Exiting the EOIs The operators remain in the EOIs until either directed out by the EOI or when the SM/US concludes that an emergency condition no longer exists. Exit from EOI-1 and associated contingency procedures always requires SM/US determination, since these procedures have no explicit exit to other plant procedures except from RC/Q to AOI-100-1. Appendix 100-1 should be reviewed prior to EOI exit to determine, restore, and document abnormal alterations that were established during EOI executio After exiting the EOIs the operator surveys the present plant conditions to ensure no reason for re-entry to the EOIs exis During Ear execution, a SAMG ENTRY IS REQUIRED condition may arise. Entry into and execution of Severe Accident Management Guidelines (SAMGs) are the responsibility ofthe SED in the TSC. Significant time may be required to man the TSC with the appropriate SAM Team members and turn over plant conditions between the control room and the TSC. The control room staff terminate execution of ALL EOI flowcharts ONLY when the SED declares that the SAM Team has assumed command and control. EOI appendices may continue in use as directed by the SAMG During the time between the development ofthe SAMG ENTRY IS REQUIRED condition and the time of assumption of command and control by the TSC, the control room staff shall continue use of available Ear guidance to mitigate the even Development of a SAMG ENTRY IS REQUIRED condition always requires entry into the SAMGs when the TSC SAM Team assumes command and control, even if plant conditions subsequently develop which seem to no longer satisfY a requirement to enter SAMGs.
) Duties of the Control Room Team Members While Executing EOls The specific duties of the Control Room Team Members are outlined in Conduct of Operation .6 Shift Communications During Execution of EOIs The methodology associated with communications during execution of the EOIs is outlined in Conduct of Operations Use ofInstrumentation and rCS/Safety Parameter Display System (SPDS) Various instruments in the control room are qualified for Post Accident Monitoring. These instruments are identified with black labels. During the performance of the EOIs, these instruments are required to be utilized as much as practical. For parameters that have multiple readouts in the control room, the operator should observe as many of the multiple readouts as practical for a verification of the values being observe Most instruments in the control room are provided with what may be considered standard scale divisions (increments of 1,5, 10, etc.), although there are some that may be considered off-normal (increments of2, 3, 4, etc.). Some pressure instruments may read out in PSIA rather than the more common value ofPSI The operator is required to remain aware of these possible differences when reading the values from the instruments. For pressure instruments, the pressure should be called out in values of PSIA or PSIG, as applicable. When the operator reading the flowchart asks for the value of a pressure parameter, it should be assumed that the value be given as PSIG unless he/she solicits the value in PSI SECTION O-VIIl-A PAGE 48 OF 52 REVISION 4
TVAN Standard Conduct of Operations OPDP-1 Department Rev. 0008 Procedure Page 56 of 103 Appendix I (Page 40f 5) Communications When using the plant PA system:
(1) Speak slowly and deliberately in a normal tone of voic (2) When announcements of abnormal or emergency conditions are made, they shall be made at least twic (3) When making announcements for drills or exercises begin and end the announcement with "This is a Drill." Plant Telephones When using Plant telephones: Identify yourself and watch statio When trying to make contact with the main Control Room, if the message is of a routine nature, the sender should hang up when the main Control Room fails to answer after the fifth ring to avoid unnecessary Control Room noise. The phone shall be allowed to ring until answered if the information is important to Operation During times when the DO NOT DISTRUB (DND) function has been used by MCR personnel, follow the directions on the recording as appropriat When making announcements for drills or exercises begin and end the announcement with "This is a Drill." Radio/phone Communication Radio/phone usage shall not be allowed in areas where electronic interference with plant equipment may resul When making announcements for drills or exercises, begin and end the announcement with "This is a Drill." Sender should identify themselves by watch statio Three way communications should be use Clear concise language should be used since radio/phone contact does not have the advantage of face to face communicatio NRC RO EXAM 75. RO GENERIC 2.4.8 001lMEM/T3111GENERIC 2.4.8//RO/SROINEW 10/16/07 Which ONE of the following describes the use of Event-Based procedures during Symptom-Based Emergency Operating Instructions (EOI) execution?
Event-Based procedures are _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ A. NOT used during Symptom-Based EOI executio B. ONLY used if specifically directed by an EOI flowchart ste C~ ONLY used if they do not interfere with EOI implementatio D. ALWAYS used if equipment or plant status require their implementatio KIA Statement: Emergency Procedures /Plan 2.4.8 Knowledge of how the event-based emergency/abnormal operating procedures are used in conjunction with the symptom-based EOP KIA Justification: This question satisfies the K/A statement by requiring the candidate to demonstrate knowledge of procedure hiearchy during execution of Emergency Operating Instruction References: EOIPM Section O-VIII-A Level of Knowledge Justification: This question is rated as MEM due to the requirement to recall or recognize discrete bits of informatio NRC Exam Friday, February 29, 2008 3:01 :07 AM 157
0610 NRC RO EXAM REFERENCE PROVIDED: None Plausibility Analysis: In order to answer this question correctly, the candidate must deterine the rules for using Event-Based procedures during EOI executio C - correct: A - incorrect: This is plausible based on the contradiction often found between Event-Based and Symptom-Based guidance. However, their use is permitted under controls circumstance incorrect: This is plausible because several EOI steps direct actions in accordance with Event-Based procedures. However, it is NOT a prerequisite to their us D - incorrect: This is plausible because no specific Event-Based procedure is expressly prohibited from use. However, if a conflict exists between the Event-Based procedure and the EOI, the EOI takes precedenc You have completed the test! Friday, February 29,20083:01 :07 AM 158
EOI PROGRAM MANUAL USER'S GUIDE FOR EMERGENCY OPERATING INSTRUCTIONS SECTiON O-Viil-A _ _ _ _ _ _ _--'-_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ EOI Flowchart Use With Other Plant Procedures The EOls are entered, based upon specific conditions symptomatic of emergencies, or conditions that could degrade into emergencies. Therefore the operator actions, provided within the EOls, allow the operator to mitigate the consequences of a broad range of accidents and multiple equipment failure Other procedures, such as AOJs, ARPs, EPIPs, etc., have event specific entry conditions and may be used to supplementEOls. In some instances the EOls will direct the operators to the unit operating procedures (Ols, GOls, and AOls) for completion of specific tasks. Usually, the EOls dir~ct the operators to specific EOIAppendices. The Appendices are specific task related procedui"es written to satisfY directives given within the EOI The exception to this rule are the SSlsand AOI-lOO-2. The conditions which cause entry into the SSls are such that the reliability of the information systems required to execute the EOls are no longer at a confidence level that would make the EOIs effectiv Any time that the operators must leave the control room, as directed by AOI-100-2, the EOls shall be exited and AOI-l 00-2 shall be used to shut down and cool down the reactor. The EO Is are not designed, or written, to support their use outside of the main control roo Conditions may arise under Station Blackout (SBO) conditions in which the rate of RPV cooldown is reduced, or alternate Heat Capacity Temperature Limit or Pressure Suppression Pressure curves are appropriate to avoid an unnecessary emergency depressurization, in order to maintain RCIC injection capability. The TSC staff or an associated abnormal operating instruction may recommend use ofthese alternate curves, which have been calculated as part ofEOIPM section 2- or 3-VI-F and -H. These alternate curves meet the assumptions used within the EOl It is recognized that during execution ofthe EOls the control room will receive assistance from various support groups. This is especially the case under conditions in the EPIPs that result in the Technical Support Center (TSC) being staffed. For example, the TSC may make recommendations regarding when it is best to vent primary containment, based upon present or predicted meteorological conditions. This would not contradict the directions provided by the EOls, but help to meet the intent of minimizing radiological releases to the general publi Execution of EOJ Appendixes The EOls rely heavily upon the EOI Appendices to implement EPG and PSTG actions and tasks that are too involved to outline on the flowchart procedure. These tasks include the defeating of various interlocks and logic systems. The steps within the Appendices involve the removing of fuses, placing jumpers across terminals, and placing boots on relay contacts, as well as some of the more common functions such as opening and closing valves and operation of systems to support the EOI flowchart procedure steps.
SECTION O-VIII-A PAGE 46 OF 52 REVISION 4 }}