ML14338A059
| ML14338A059 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 12/03/2014 |
| From: | NRC/RGN-II |
| To: | |
| References | |
| Download: ML14338A059 (242) | |
Text
PrefixList:A.0-Question A.1-DistractorAnalysis A.2-MiscComments/Feedback/History A.3-NUREGES 401 5 A.4-OriginalQuestion(ifquestionwasMODIFIEDfromaBANKquestion) A.5-SuppliedReferenceB.0-RegulatorDocuments B.0-TechSpecs B.1-TechSpecBases B.2-TRM B.3-COLR B.4-PTLR B.5-ODCM B.6-PLS B.7-PTDBC.0-Procedures C.1-EOP C.2-AOP C.3-SOP C.4-UOP C.5-ARP C.6-Admin C.9-OtherD.0-Drawings D.1-P&IDs D.2-Oneline D.3-Elementary D.4-Logic D.5OtherE.0-MiscOther E.1-Photographs E.2-Maps E.3-RadSurveys E.4-LessonPlan
- 1. 001AA2.04 001/LOIT/SRO/C/A 4.2/4.3/001AA2.04/LO-TA-63013/// Initial conditions:
- Unit 1 is at 70% reactor power. - Main Turbine load is stable.
- Rods are in aut omatic with CBD at 190 steps.
- Control rods star t stepping out without demand.
Current conditions:
- Rod motion stops when the Rod Bank Selector Switch is placed in Manual.
- Reactor power indication has risen to 75%.
- CBD rods are at 211 st eps, EXCEPT control rod H8, which is at 190 steps by DRPI indication.
With NO other actions taken, which one of the following completes the following statement?
When the inadvertent rod motion stops , reactor power indication will __(1)__, and per the Bases of Tech Spec 3.1.4, "Rod Group Alignment Limits," CBD rod H8 is__(2)__ at this time.
(1) stabilize and remain near 75%
(2) OPERABLE (1) stabilize and remain near 75%
(2) inoperable (1) trend down from the peak value observed to slightly above 70%
(2) OPERABLE (1) trend down from the peak value observed to slightly above 70%
(2) inoperable A.B.C.D.K/A001 Continuous Rod Withdrawal AA2.04 Ability to determine and interpret the following as they apply to theContinuous Rod Withdrawal: - Reactor power and its trendThursday, February 20, 2014 8:28:29 AM 1
K/A MATCH ANALYSIS The question sets up a plausible scenario where rods withdrawal due to some internal failure inserts some amount of positive reactivity. Once operator action is taken rod motion stops and the candidate must address how the core will respond to the inserted reactivity and the impact of the stuck rod, ther efore the two element s are in place that the KA requires. The OPERABILITY determination for the affect rod is SRO required knowledge.EXPLANATION OF REQUIRED KNOWLEDGE As control rods are withdrawn, an increase in fission rate will occur. The resulting fuel centerline temperature increase from the positive reac tivity addition will be offset by FTC and MTC feedback. As a result, reactor power will tr end down toward the orginial power level with and increas e RCS average temper ature. Since RCS temperature is higher than the orginial temperature, SG pr essure will also be slig htly elevated. The higher SG pressure will result in slight increase in steam flow as compared to the original conditions. Ther efore, reactor power will be slightly higher than thepre-transient value.
This is a fundamental reactor theory concept.
Per TS 3.1.4 Bases, the OPERABILITY requir ements (i.e. trippability) are seperate from the alignment requirem ents. Even if a rod is >12 steps misaligned, it is still OPERABLE.
Where rod(s) are not moving , the rod(s) must be consider ed untrippable unless there isverification that a rod control system failure is preventing rod motion. Since rod motion was demanded and did not occur and there is not hing given in the st em to explain the loss of motion, the candidate must declare rod H8 inoperable.ANSWER / DISTRACTOR ANALYSISA. Incorrect. Plausible. Part 1 is 'plausib le' however incorrect in that the candidate may determine that reactor power will rise to a new higher equilibrium value and stabilize.
Thus, the candidate failed to take into count the reactivity feedback mechanisms of MTC and FTC on core behavior.
In addition, the answer would be correct under some plant condition like during startup steam dumps in steam pressure mode. Part 2 is 'plausible' but is also incorrect in that the candidate must consider the affected r od inoperable until proven trippable as stated in Tech Spec 3.1.
4 'Rod Group Alignment Limits' bases; However, where rod(s) are not moving, the rod(s) must be considered untrippabl e unless there is veri fication that a rodcontrol system failure is preventing rod motion. If the rod controlsystem is demanding motion properly and no motion occurs, the
rod is considered untrippable (i.e., inoperable). This operability call is contrary to the standard philosophy that a component is
considered OPERABLE until determined otherwise.B. Incorrect. Plausible. Part 1 is 'plausib le' however incorrect. See Part 1 of choice A above.Thursday, February 20, 2014 8:28:29 AM 2
Part 2 is correct requiring the candidate to recall Tech Spec 3.1.4 'Rod Group Alignment Limits' bases which tiesOPERABILITY to trippability of the rods. The rod OPERABILITY(i.e., trippability) requirement is satisfied provided that the rod will fully insert in the require d rod drop time assumed in the safety analyses. Rod control malf unctions that result in the inability to move a rod (e.g., rod lift coil failures), but that do not impact trippability, do not result in rod inoperability. However, where rod(s) are not moving, t he rod(s) must be considereduntrippable unless there is verification that a rod control system failure is preventing rod motion. If the rod control system is
demanding motion properly and no motion occurs, the rod is
considered untrippable (i.e., inoperable).
C. Incorrect. Plausible. Part 1 is corr ect which is addressing core response and correctly predicts that the positive reactivity will be offset by FTC and MTC feedback. Over time the reactor power will lower to near original values, onl y slightly higher due to T AVG being raised which results in an increase in Steam Generator pressure. This would result in a small increase in steam flow. Part 2 is 'plausible' but is also incorrect. See Part 2 of choice A above.D. Correct. Part 1 is correct.
See Part 1 of choice C above.
Part 2 is correct. See Part 2 of choice B above.SRO JUSTIFICATION (10CFR43(b))(2) Facility operating limitations in the technical specifications and their bases.-Can question be answered solely by knowing = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action?
No thequestion is not addressing any TS action times.-Can question be answered solely by knowing the LCO/TRM information listed "above-the-line?"
No, the question is not addressing above-the-line TSinformation. The required knowledge is TS Bases.-Can question be answered solely by knowing the TS Safety Limits?
No, thequestion is not related to TS Safety Limits.-Does the question involve one or more of the following for TS,TRM, or ODCM?
- Application of Required Actions (Sec tion 3) and Surve illance Requirements (Section 4) in accordance with rules of applicati on requirements (Section 1)* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)* Knowledge of TS bases that is required to analyze TS required actionsand terminology.
Yes. Specific knowledge of TS Bases is required tomake the OPERABILITY call required an d this call is contrary to thestandard philosophy for OPERABILITY calls.Thursday, February 20, 2014 8:28:29 AM 3
Level: SROTier # / Group # T1 / G2 K/A# 001AA2.04 Importance Rating: 4.2 / 4.3 Technical
Reference:
Tech Spec 3.1.4 Bases Rev 1-8/03References provided: None Learning Objective: LO-LP-36990-07 State the Techni cal Specification basesfor the restrictions on control rod insertion
limit, and alignment (SRO only).
LO-LP-39205-02 Given a set of Tech Specs and the Bases, determine for a specific set of plant
conditions, equipment availability, and operational mode: Whether any Tech
Spec LCO's of section 3.1 are
exceeded.The required actions for all section 3.1. LCO's. LO-TA-63013 Implement Tec hnical Specification LCO using 10008-C (SRO Only)Question origin: MODIFI ED - Vogtle HL18 Q uestion #005AG2.1.07 001Cognitive Level: C/A 10 CFR Part 55 Content: 41.1 / 43.2 Comments:
You have completed the test!Thursday, February 20, 2014 8:28:29 AM 4
- 1. 005AG2.1.07 001/1/2/STUCK CRDM/C/A - 4.4/4.7/BANK-HL-17/HL-18 NRC/SRO/AML Initial conditions:
- Time = 0900.
- Unit 1 is at 60% power following a refueling outage.
- The OATC is withdrawi ng rods when one DRPI is se en not moving with its group. - The OATC immediatel y stops withdrawing rods, and all rod motion stops.
- CBD, Group 2, Rod H-8 DRPI indicates 198 steps.
- CBD, Group 2, step counters indicate 209 steps.
Current conditions:
- Time = 0945.
- No rod motion has occurred since 0900.
- I&C has verified no faults on the DRPI system.
- I&C has verified that the rod li ft coil for Control Rod H-8 is failed.
Which one of the following comp letes the below statements?
Based on the initial conditions, at 0900 Contro l Rod H-8 was __________ in accordance with the Bases of Tech Spec 3.1.4, Rod Group Alignment Limits.
Based on the current conditions, at 0945 Control Rod H-8 was __________ in accordance with the Bases of Tech Spec 3.1.4, Rod Group Alignment Limits. Rod H-8 status Rod H-8 Status at 0900 at 0945 OPERABLE inoperable inoperable inoperable OPERABLE OPERABLE inoperable OPERABLE A.B.C.D.Monday, January 20, 2014 9:38:55 AM 1
Rod Group Alignment Limits B 3.1.4 (continued)Vogtle Units 1 and 2 B 3.1.4-5 Rev. 1-8/03 BASES (continued) LCO The limits on shutdown or control rod alignments ensure that the assumptions in the safety analysis will remain valid. The requirements on OPERABILITY ensure that upon reactor trip, the assumed reactivity will be available and will be inserted. The OPERABILITY requirements (i.e., trippability) are separate from the alignment requirements which ensure that the RCCAs and banks maintain the correct power distribution and rod alignment.
The rod OPERABILITY (i.e., trippability) requirement is satisfied provided that the rod will fully insert in the required rod drop time assumed in the safety analyses. Rod control malfunctions that result in the inability to move a rod (e.g., rod lift coil failures), but that do not impact trippability, do not result in rod inoperability. However, where rod(s) are not moving, the rod(s) must be considered untrippable unless there is verification that a rod control system failure is preventing rod motion. If the rod control system is demanding motion properly and no motion occurs, the rod is considered untrippable (i.e.,
inoperable). The requirement to maintain the rod alignment to within plus or minus 12 steps of their group step counter demand position is conservative. The safety analysis assumes a total misalignment from fully withdrawn to fully inserted. When required, movable incore detectors may be used to determine rod position and verify the rod alignment requirement of this LCO is met. Failure to meet the requirements of this LCO may produce unacceptable power peaking factors and LHRs, or unacceptable SDMs, all of which may constitute initial conditions inconsistent with the safety analysis. APPLICABILITY The requirements on RCCA OPERABILITY and alignment are applicable in MODES 1 and 2 because these are the only MODES in which a self-sustaining chain reaction (K eff 1) occurs, and the OPERABILITY (i.e., trippability) and alignment of rods have the potential to affect the safety of the plant. In MODES 3, 4, 5, and 6, the alignment limits do not apply because the control rods are fully inserted and the reactor is shut down, with no self-sustaining chain reaction. In the shutdown MODES, the OPERABILITY of the shutdown and control rods has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the RCS. See LCO 3.1.1, "SHUTDOWN MARGIN (SDM)," for SDM in MODES 3, 4, and 5 and LCO 3.9.1, "Boron Concentration," for boron concentration requirements during refueling.
- 1. 001G2.1.19 001/LOIT/SRO/C/A 3.9/3.8/001G2.1.19/LO-TA-05002///G2.1.37 Initial conditions:
- Unit 1 reactor power lower ed due to an inadver tent turbine runback. - ALB10-D04 ROD BANK LO-LO LIMIT alarm was received.
Current conditions:
- Main Turbine load has been stabilized.
- RCS Tavg is 2°F below Tref.
- The OATC has requested a 2-step rod withdraw al for temperature control.
Which one of the following completes the following statement?
Based on the current conditions and usin g the Plant Computer data provided, ALB10-D04 __(1)__ valid, and based on the current conditi ons and NMP-OS-001, "React ivity Management Program," guidance, the Shift Supervisor is
__(2)__ to authorize a 2-step control rod withdrawalfor Tavg control.REFERENCE PROVIDED __(1)__ __(2)__ is allowed is NOT allowed is NOT allowed is NOT NOT allowed A.B.C.D.K/A 001 Control Rod DriveG2.1.19 Ability to use plant computers to evaluate system or component status.K/A MATCH ANALYSIS The question test the candidat e ability to evaluate and valida te a plant computer (IPC) generated alarm for the current plant co nditions associated with the Rod Controlsystem. Then the SRO is required to determine if control rod withdrawal is both prudent and allowed under the stat ed conditions in the stem pe r the NMP-OS-001 'Reactivity Management Program'.Thursday, February 20, 2014 8:45:05 AM 1
EXPLANATION OF REQUIRED KNOWLEDGE The IPC calculates the rod insertion limit using real time data. When the alarm is generated the candidate would be expected to verify the va lidity of the alarm. This evaluation would be accomplished using IPC and Te ch Spec data with backup from QMCB indications. In addition, the Shift Supervisor is r equired the make a decision to withdrawal control r ods based on NMP-OS-001 'R eactivity Management Program' guidance.The control room team shall not immediatel y dilute or withdraw control rods in an attempt to restore RCS Tavg
/Tref deviations caused by a secondary plant transient.Attempts to immediately restore RCS Tavg/T ref deviations caused by a secondary plant transient can be aggravated by withdra wing control rods or reducing boron concentration with reactor power rising.
Per NMP-OS-001, once tu rbine load has been stabilized and RCS Tavg has been restored to wit hin 3°F of Tref, positive reactivity can be added by withdra wing control rods.ANSWER / DISTRACTOR ANALYSISA. Correct. The first part is correct. Per the IPC printout, RCS DeltaT Power is approximately 83% and CBD position is 116 steps.
The IPC uses Auctioneered Hi gh DeltaT Power for the RIL calculation. Per the COLR, th e RIL for 83% po wer is 122 steps on CBD. Therefore, ALB10-D04 is a valid alarm.
The second part is correct.
Per NMP-OS-001 step 6.1.2.4, once turbine load has been stab ilized and RCS Tavg has been restored to within 3° of Tref, po sitive reactivity can be added by withdrawing control rods. Both of these conditions have been met. Therefore, control r od withdrawal is not restricted.B. Incorrect. Plausible. The first part is correct. See the first part of Choice A above.
The second part is incorrect.
Per NMP-OS-001 step 6.1.2.4, rod withdrawals are allowed if Tavg/Tref devation is within 3°F.
However, NMP-OS-001 st ep 6.1.2.4 st ates it is non-conservative to wi thdraw control rods in response to a transient and anomalies are to be mitigated utilizing the
secondary plant. It is reaso nable for a candidat e not familiar with the specific guidance of NM P-OS-001 to believe that the Tavg/Tref of 2°F is still consi dered a "transient" and not allow the use of control rods. Therefor e, this distractor is plausible.
C. Incorrect. Plausible. The first part is incorrect. Per the IPC printout, RCS DeltaT Power is approximately 83% and CBD position is 116 steps.
Per the COLR, the RIL for 83% power is 122 steps on CBD.
Therefore, ALB10-D04 is a valid alarm. However, if the candidate does not understand ho w RIL is calculated and usesNIS Power instead of DeltaT Po wer, an RIL of 105 steps would be determined and the candidate w ould conclude the alarm wasThursday, February 20, 2014 8:45:05 AM 2
not valid. Therefor e, this distractor is plausible.
The second part is correct. See the second part of choice B above.D. Incorrect. Plausible.
The first part is incorrect. See the first part of choice C above.
The second part is correct. See the second part of choice B above.SRO JUSTIFICATION (10CFR43(b))(5) Assessment of facility conditions and selection of appropriate proceduresduring normal, abnormal, and emergency situations.-Can the question be answered solely by knowing "systems knowledge", i.e.,how the system works, flowpath, logic, component location?
No, specificknowledge of Reactivity management during a transient per NMP-OS-001 isrequired.-Can the question be answered solely by knowing immediateoperator actions?
No, there are not associated IOA's.-Can the question be answered solely by knowing entry conditions for AOPs orplant parameters that require direct entry to major EOPs?
No, the only associated AOP (18013-C) does not directly address the RIL alarm or theassociated temperature requirement.-Can the question be answered solely by knowing the purpose, overall sequenceof events, or overall mitigative strategy of a procedure?
No, the conditions are specific to NMP-OS-001.-Does the question re quire one or more of the following?* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or se ction of a procedure to mi tigate, recover, or with which to proceed* Knowledge of when to impl ement attachments and appendices, including how to coordinate these it ems with procedure steps* Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific s ubprocedures or emergency contingency procedures* Knowledge of administrative procedures that specify hierarchy,implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
Yes, the question requires specific knowledge ofNMP-OS-001, which is an administrative procedure on Reactivity Management and more specifically the guidance associated withmanagement of a transient condit ion and the duti es of the SRO(Reactivity Management SRO) when reactivity manipulations are beingperformed.(6) Procedures and limitations involved in initial core loading, alternations incore configuration, control rod programming, and determination of variousinternal and external effects on core reactivity.Thursday, February 20, 2014 8:45:05 AM 3
The question is about "Procedures and limitations involved in control rod movement as it relates to internal/external effects on core reactivity".
Although not listed as an example for this category in the "ClarificationGuidance for SRO-only Questions", this question does include proceduraladministrative requirements and controls associated with external effects oncore reactivity.
Level: SROTier # / Group # T2 / G2 K/A# 001G2.1.19 Importance Rating: 3.9 / 3.8Technical
Reference:
NMP-OS-001 Rev 17.0, page 10 Unit 1 Cycle 18 COR, Figure 3
ARP 17010-1 Rev 50, page 3 and 41References provided: Plant Computer RIL screenshot and COLR Figure 3Learning Objective: LO-LP-39205-07 State th e reasons for maintaining rods above the RIL.
LO-LP-60301-08 Descr ibe how placing the delta T defeat switch to a failed channel will affect the
response of the rod insertion limit
computer.LO-PP-27101-21 Stat e the alarms associated with the rod insertion limits; in clude set points and source of the set points.LO-TA-05002 Obtain Data Fr om the Integrated Plant Computer using 13505-1/2Question origin: BANK - Hatc h 2011 NRC Question # G2.1.37Cognitive Level: C/A 10 CFR Part 55 Content: 43.5 Comments: Early submittal 401-9 response:
Question appears to match the KA. Question appearsto be at the SRO level.
Question appears to be okayas-is.
- JAT 12/19/13 (SAT)You have completed the test!Thursday, February 20, 2014 8:45:05 AM 4
CORE OPERATING LIMITS REPORT, VEGP UNIT 1 CYCLE 18JULY 2012 Page 9 of 11 FIGURE 3ROD BANK INSERTION LIMITS VERSUS % OF RATED THERMAL POWER POWER (% of Rated Thermal Power)ROD BANK POSITION (Steps Withdrawn)
BANK D BANK C BANK B0102030405060708090100 0 20 40 60 80100120140160 180200220(30.2%, 0)(100%, 161)(78.0%, 225)(0%, 161)(28.0%, 225)(Fully Withdrawn
- )*Fully withdrawn shall be the condition where control rods are at a position within the interval 225 and 231 steps withdrawn.
NOTE: The Rod Bank Insertion Limits are based on the control bank withdrawalsequence A, B, C, D and a control bank tip-to-tip distance of 115 steps.(0%, 46)
CORE OPERATING LIMITS REPORT, VEGP UNIT 1 CYCLE 18JULY 2012 Page 9 of 11 FIGURE 3ROD BANK INSERTION LIMITS VERSUS % OF RATED THERMAL POWER POWER (% of Rated Thermal Power)ROD BANK POSITION (Steps Withdrawn)
BANK D BANK C BANK B0102030405060708090100 0 20 40 60 80100120140160 180200220(30.2%, 0)(100%, 161)(78.0%, 225)(0%, 161)(28.0%, 225)(Fully Withdrawn
- )*Fully withdrawn shall be the condition where control rods are at a position within the interval 225 and 231 steps withdrawn.
NOTE: The Rod Bank Insertion Limits are based on the control bank withdrawalsequence A, B, C, D and a control bank tip-to-tip distance of 115 steps.(0%, 46)
Approved By J.B. Stanley Vogtle Electric Generating Plant Procedure Number Rev 17010-1 50 Date Approved 08/16/2011 ANNUNCIATOR RESPONSE PROCE DURES FOR ALB 10 ON PANEL 1C1 ON MCB Page Number 3 of 66 Printed September 26, 2013 at 11:48 ALB 10 (1) (2) (3) (4) (5) (6)
A SR/IR SIG PROCESSOR
TROUBLE
NIS SOURCE AND INTMD RANGE TRIP BYPASS
POWER RANGE HI NEUTRON FLX HI
SETPOINT ALERT
REACTOR BYPASS BRKR BYA IN-OPERATE
REACTOR BYPASS BRKR BYA CLOSE
ROD CONTROL
NON URGENT FAILURE B
SOURCE RNG HI SHUTDOWN FLUX
ALARM BLOCKED
POWER RANGE HI NEUTRON FLX LOW SETPOINT
REACTOR BYPASS BRKR BYB IN-OPERATE
REACTOR BYPASS BRKR BYB CLOSE ROD CONTROL URGENT FAILURE
C SOURCE RANGE HI FLUX LEVEL AT SHUTDOWN
POWER RANGE
CHANNEL DEVIATION OVERPOWER T ROD BLOCK AND
RUNBACK ALERT
ROD BANK LO LIMIT
RPI NON URGENT
ALARM NIS CHANNEL
ON TEST
D INTMD RANGE HI FLUX LEVEL ROD STOP
PWR RANGE UP DET HI FLX DEV
OVERPOWER
ROD STOP ROD BANK LO-LO LIMIT
RPI URGENT ALARM
ROD DEV E
SR/IR REMOTE
SIG PROCESSOR
DPU-B TROUBLE
OVERTEMP T ROD BLOCK AND
RUNBACK ALERT
ROD AT BOTTOM
RADIAL TILT
F SR/IR AMPLIFIER
TROUBLE
POWER RANGE HI NEUTRON FLX
RATE ALERT
ROD DRIVE M-G
SET TROUBLE
TWO OR MORE RODS AT BOTTOM
DELTA FLUX
DEVIATION
Approved By J.B. Stanley Vogtle Electric Generating Plant Procedure Number Rev 17010-1 50 Date Approved 08/16/2011 ANNUNCIATOR RESPONSE PROCE DURES FOR ALB 10 ON PANEL 1C1 ON MCB Page Number 41 of 66 Printed September 26, 2013 at 11:48 WINDOW D04 ORIGIN SETPOINT IPC Calculated UD0366 Rod Insertion Limit
ROD BANK LO-LO LIMIT
1.0 PROBABLE
CAUSE RCS Boron concentration too low for present reactor power level due to:
- 1. Plant transient.
- 3. IPC failure
2.0 AUTOMATIC
ACTIONS NONE
3.0 INITIAL
OPERATOR ACTIONS
- 1. Check indications and determine if actual control bank rod position is below the Lo-Lo insertion limit by referring to the COLR and Technical
Specification LCO 3.1.6.
- 2. IF actual control bank position is below the Lo-Lo Insertion Limit, perform the following:
- a. Within 1 hour:
Verify shutdown margin is within the limits specified in the COLR per 14005-1 "Shutdown Margin Calculation";
Refer To TR 13.1.1 for applicability.
OR Initiate and maintain Emergency Boration per 13009-1, "CVCS Reactor Makeup Control System", until the Control Banks Lo-Lo
Limit Annunciator clears.
- b. Restore the affected control bank(s) ab ove the limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Southern Nuclear Operating Company Nuclear Management Procedure Reactivity Management Program NMP-OS-001 Version 17.0 Page 10 of 39 6.1.2.3 Administrative Controls The following administrative requirements ensure that nuclear safety is maintained during activities that affect reactivity:
Planned reactivity manipulations are peer checked.
Calculations that involve reactivity control are independently verified prior to use.
Reactor operators use redundant instrumentation when monitoring the effects of reactivity manipulations.
Reactor engineering is actively engaged in activities that change reactivity significantly, including any special tests with the potential to affect reactivity.
The reactor operator performing rod movement activities is free from distractions and will have no other duties while performing reactivity manipulations.
See Attachment 2 for summarized expectations for reactivity manipulations.
6.1.2.4 Conduct of Reactivity Changes Operators anticipate the effects of reactivity manipulations and monitor core parameters carefully until parameters stabilize. Any unanticipated reactivity change is immediately brought to the attention of management and the resolution of the change is pursued to its conclusion.
Adding positive reactivity is never an appropriate way to address unstable plant conditions. It is non-conservative to withdraw control rods in response to primary plant anomalies caused by unplanned secondary plant transients. For the PWRs, once turbine load has been stabilized and RCS Tavg has been restored to within 3 degrees of Tref, positive reactivity can be added by withdrawing control rods.
Whenever the status of reactor criticality becomes unknown, the reactor is shutdown.
Under normal conditions, positive reactivity changes will not be performed by more than one means at a time.
During approach to criticality, two positive reactivity additions will not be performed
simultaneously.
The operator at the controls shall suspend turnover if a reactivity manipulation is
required during turnover.
- 1. 003G2.4.3 001/LOIT AND LOCT/SRO/M/F 3.7/3.9/003G2.4.3/LO-TA-37015///
Initial conditions:
- Unit 1 experienced a small break LOCA.
- 19000-C, "Reacto r Trip or Safety Injection," is in progress.
- Loop 3 RCS Tcold in strumentation is not available.
Current condition:
- RCS pressure is 1350 psig and slowly lowering.
Which one of the following completes the following statement?
Stopping the RCPs is required to minimize the risk of __(1)__, and per the Bases of Tech Spec 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," for diverse indication of RCS Tc old temperature, t he operator is dire cted to use __(2)__
instrumentation. __(1)__ __(2)__ core uncovery RCS Thot core uncovery steam generator pressure RCP damage RCS Thot RCP damage steam generator pressure A.B.C.D.K/A 003 Reactor Coolant PumpG2.3.4 Ability to identify post-accident instrumentation.K/A MATCH ANALYSIS The question tests the candidate's ability to identify the post-acci dent instrument that would be utilized as a diverse indication of RCS Tcold temperature with RCPs secured.EXPLANATION OF REQUIRED KNOWLEDGE Per WOG Background for RCP Trip, the reason for purpo sely tripping the RCPs during a small break LOCA is to pr event excessive depletion of RCS water invent ory through a saml break in the RCS which might lead to severe core uncover y if the RCPs were tripped for some reason later in the acci dent. The RCPs should be tripped before the RCS inventroy is depleted to the point where tripping of the pumps would cause theThursday, February 20, 2014 1:26:28 PM 1
break to immediately uncover. The WOG gives options in par ameter that may be utilzied as RCP trip criteria. Vogtle has chosen the option based solely on RCS pressure. Therefore, RCP s are manually tripped if RC S pressure is <1375 psig provided either CCPs or SIPs are injecting into the core.
Per TS 3.3.3 FU 2,3 Bases, st eam line pressure provides diverse indication for the RCS cold leg temperature. With either forced or na tural circulation flow through the steam generators, SGs will be at saturation pressure for the RCS Cold Legs.ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. Part 1 of the answer is correct and states the reason provided inthe Westinghouse background documents for tripping the
Reactor Coolant Pumps for small break LOCAs as the potential for core damage due to core uncover and exceeding peek centerline temperat ures criteria.
Part 2 is not correct but is 'plausible' because T-hot instrumentation is addressed in the same bases as diverse indication for the CETCs as opposed to Tcold instrumentation.
The candidates would see there is a relationship since both are measuring temperature as oppos ed to the correct instrument which is using pressure.B. Correct. Part 1 is correct.
See Part 1 of choice A above.Part 2 is correct. Per Tech Spec 3.3.3 'Post Accident Monitoring Instrumentation,' st eam line pressure provides diverse indication for the RCS cold leg temperature. C. Incorrect. Plausible. Part 1 is not corr ect for the small break LOCA but would be true for large break LOCA s since the Reactor Coolant Pumps would be stopped under this conditio n due to loss of support conditions for continued oper ation and subsequent pump damage.Part 2 is not correct. See Part 2 of choice A above.D. Incorrect. Plausible. Part 1 is not correct. See Part 1 of choice C above.
Part 2 of the answer is correct.
See Part 2 of choice B above.SRO JUSTIFICATION (10CFR43(b))(2) Facility operating limitations in the technical specifications and their bases.-Can question be answered solely by knowing = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action?
No, thequestion is not addressing Tech Spec action times.-Can question be answered solely by knowing the LCO/TRM information listed "above-the-line?"
No, the question is not addr essing Tech Spec above-the-lineinformation.Thursday, February 20, 2014 1:26:29 PM 2
-Can question be answered solely by knowing the TS Safety Limits?
No, thequestion is not related to Tech Spec Safety Limits.-Does the question involve one or more of the following for TS,TRM, or ODCM?
- Application of Required Actions (Sec tion 3) and Surve illance Requirements (Section 4) in accordance with rules of applicati on requirements (Section 1)
- Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)* Knowledge of TS bases that is required to analyze TS required actionsand terminology.
Yes, the answer to the question is only found in Tech Spec bases.
Level: SROTier # / Group # T2 / G1 K/A# 003G2.4.3 Importance Rating: 3.7 / 3.9Technical
Reference:
Westinghouse Ba ckground - RCP Trip Rev 2, 4/30/2005 Tech Spec 3.3.3 Bases page B 3.3.3-6, Rev 0Reference Provided: None Learning Objective: LO-TA-37006 Conduc t a Natural Circulation Cooldown per 19002-CLO-TA-37015 Perform the Initia l Recovery Actions for a small Loss of Reactor or Secondary
Coolant per 19010-CLO-TA-63013 Implement Tec hnical Specification LCO using 10008-C (SRO Only)
LO-LP-39207-04 Descr ibe the bases for any given Tech Spec in section 3.3.
LO-LP-39208-01 For any given it em in section 3.4 of Tech Specs, be able to: St ate the LCO. State any one hour or less required actions.Question origin: NEW Cognitive Level: M/F
10 CFR Part 55 Content: 41.10 / 43.2Comments: You have completed the test!Thursday, February 20, 2014 1:26:29 PM 3
PAM Instrumentation B 3.3.3 Vogtle Units 1 and 2 B 3.3.3-6 Revision No. 0 BASESLCO 2,3. Reactor Coolant System (RCS) Hot and Cold Leg Temperatures (Wide Range) (continued) RCS hot and cold leg temperatures are used to determine RCS subcooling margin. RCS subcooling margin will allow termination of safety injection (SI), if still in progress, or reinitiation of SI if it has been stopped. RCS subcooling margin is also used for unit stabilization and cooldown control. In addition, RCS cold leg temperature is used in conjunction with RCS hot leg temperature to verify the unit conditions necessary to establish natural circulation in the RCS. Reactor outlet temperature inputs to the Reactor Protection System are provided by two fast response resistance elements and associated transmitters in each loop. The channels provide indication over a range of 50F to 700F. The core exit thermocouples provide diverse indication for the RCS hot leg temperature. Steam line pressure provides diverse indication for the RCS cold leg temperature. 4. Steam Generator Water Level (Wide Range) Wide range SG water level (Loops 501, 502, 503, & 504) is a Type A variable used to determine if an adequate heat sink is being maintained through the SGs for decay heat removal, primarily for the response to a loss of secondary heat sink event when the level is below the narrow range. The wide range SG level indication may also be used in conjunction with auxiliary feedwater flow for SI termination. In addition, the wide range level is cold calibrated and provides a complete range for monitoring SG level during a cooldown. Auxiliary feedwater flow provides the diverse indication for wide range SG water level. (continued)
~~
~~
~
$!$'(
- 1. 004G2.2.44 001/LOIT AND LOCT/SRO/C/A 4.2/4.4/004G2.2.44/LO-TA-16009///
Given the following procedure titles:
- 19000-C, "Reactor Trip or Safe ty Injection" - 13003-1, "Reacto r Coolant Pump Operation" - 18005-C, "P artial Loss of Flow" Initial condition:
- Unit 1 is at 16% reac tor power with a startup in progress.
Current conditions:
- ALB08-A04 RCP 1 NO. 2 SEAL LKOF HI FLOW is received. - ALB08-A05 RCP 1 CONTROLLED LKG HI/LO FLOW is received.
- 1FI-160A, #1 SEAL LEAK-OFF for RCP #1 is indicating 6.0 gpm.
Which one of the following completes the following statement?
RCP #1, seal number __(1)__ has failed, and per 13003-1, the Shift Supervisor will direct the crew to __(2)__. __(1)__ __(2)__ one initiate 18005-C one trip the reactor and initiate 19000-C two initiate 18005-C two trip the reactor and initiate 19000-C A.B.C.D.K/A 004 Chemical and Volume ControlG2.2.44 Ability to interpret control room indications to verify the status andoperation of a system, and understand how operator actions and directives affect plant and system conditions.K/A MATCH ANALYSIS The question tests the candid ate's ability to interpret t he control room annunciators associated with RCP seal #1 and
- 2 leak-offs. Seal injectio n and leakoff are part of theCVCS system per the K/A catalog. The candidate is then required to utilize these annunciators to diagnose th e issue and select the appr opriate procedure path toTuesday, February 25, 2014 8:43:22 AM 1
address the degraded conditions.EXPLANATION OF REQUIRED KNOWLEDGE Per ARP 17008-1 window A05, either low or high seal flow can actuate the alarm. High seal flow is an indication of Seal #1 failure, and low seal flow is an indication of Seal #2 failure. The candidate is given a seal flow of 6.0 gpm to allow differentiation between these two. Window A04 list s a probable cause of Seal #2 failure only. Knowledge of RCP seal package constructi on is required to understa nd why this annunciator would be consistent with a Seal #1 failure. As Seal #1 fails, the seal surfaces open and backpressure lowers within the package. Fl ow down the shaft lowers and increases both to the leak-off path and to Seal #2 .
Per 13003-1, Seal #1 leak-off >5.5 gpm requires stopping the RCP immediately per step 4.2.1.4. With reactor power greater than 15%
RTP, the operator is directed to trip the reactor and initiate 19000-C. When the IOA
's are complete, the operator is direct to perform steps 4.2.1.4.d thru h to stop the RCP, close the associated spray valve, and isolate seal #1 leak-off.
The question stem specifies 1FI-160A, #1 SEAL LE AK-OFF for RCP #1 is indicating 6.0 gpm. This is top of scale for the indicator. Simula tions show that a seal failure equivalent to approximately 9 gpm leak leak-off is required for annu nciators ALB08-A04 RCP 1 NO. 2 SEAL LKOF HI FLOW an d ALB08-A05 RCP 1 CONTROLLED LKG HI/LO FLOW to be in alarm. At a seal leak-off of >4.8 gpm and
<9 gpm, only annunciator ALB08-A05 RCP 1 CONTROLLED LKG HI/LO FLOW is in alarm.ANSWER / DISTRACTOR ANALYSISA. Incorrect. Plausible. The first part is correct. The combinatio n of annunciator alarms and elevated seal #1 leak-off are symptoms of a seal #1 failure.
The second part is inco rrect. With seal
- 1 leak-off > 5.5 gpm, an immediate shutdown of the RCP is required. Per 13003-1 step 4.2.1.4, the reacto r must be tripped and 19000-C initiated.
However, if a candidate does not realize RTP is >15%, then
step 4.2.1.4 would direct initiation of AOP 18005-C.B. Correct. The first part is correct. See the first part of choice A above.
The second part is correct. With seal #1 leak-off > 5.5 gpm, an immediate shutdown of the RCP is required. Per 13003-1 step 4.2.1.4, the reactor must be tripped and 19000-C initiated.
C. Incorrect. Plausible. The first part is incorrect. The comb ination of annunciator alarms and elevated seal #1 leak-off are symptoms of a seal #1 failure. However, if the candidate does not recognize 1FI-160A
indicating 6.0 gpm as being abn ormally high, then annunciator ALB08-A04 RCP 1 NO. 2 SEAL LKOF HI FLOW would be a symptom of a seal #2 failure.
The second part is inco rrect. With seal
- 1 leak-off > 5.5 gpm,Tuesday, February 25, 2014 8:43:22 AM 2
an immediate shutdown of the RCP is required. Per 13003-1 step 4.2.1.4, the reacto r must be tripped and 19000-C initiated.
However, if a candidate has diagnosed a seal #2 failure, 13003-C still supports an immedi ate stop of the RCP given other conditions. Si nce the question does not give an option to leave the RCP running, the c andidate must assume a valid reason to stop the RCP has been met. As such, if the candidate then does not realize RTP is >15%, then step 4.2.1.4 would direct initiation of AOP 18005-C.
D. Incorrect. Plausible. The first part is in correct. See the first part of choice C above.
The second part is correct. See the second part of choice B above.SRO JUSTIFICATION (10CFR43(b))(5) Assessment of facility conditions and selection of appropriate proceduresduring normal, abnormal, and emergency situations.-Can the question be answered solely by knowing "systems knowledge", i.e.,how the system works, flowpath, logic, component location?
No, the first part issystem knowledge however, the second part is discriminating to the SRO level asit requires an op erational decision.-Can the question be answered solely by knowing immediate operator actions? No, the requrired actions are specific direction associated with ARPs and SOPsand a shutdown decision must be made.
-Can the question be answered solely by knowing entry conditions for AOPs orplant parameters that require direct entry to major EOPs? No, entry conditions to18005-C and 19000-C both addres s plant conditions in a generic nature. Specificprocedure knowledge is required to differentiate the required procedure flow path.
-Can the question be answered solely by knowing the purpose, overall sequenceof events, or overall mitigati ve strategy of a procedure?
No, the questionrequires specific knowledg e of a specific SOP step.-Does the question re quire one or more of the following?* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or se ction of a procedure to mi tigate, recover, or with which to proceed.
Yes, the candidate must use specific knowledge of SOP13003-C decision flow charts from memory as well as specific knowledgeof step 4.2.1.4 in order to direct the specific actions to be taken toshutdown the reactor and a ddress the RCP seal failure.* Knowledge of when to impl ement attachments and appendices, including how to coordinate these it ems with procedure steps* Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific s ubprocedures or emergency contingency procedures* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant norma l, abnormal, and emergency proceduresTuesday, February 25, 2014 8:43:22 AM 3
Level: SROTier # / Group # T2 / G1 K/A# 004G2.2.44 Importance Rating: 4.2 / 4.4Technical
Reference:
ARP 17 008-1, Rev 18.0, pages 10-12 SOP 13003-1, Rev 47.1, pages 13-15 & 38-40 AOP 18005-C, Rev 11.1, pages 1 & 3
V-LO-PP-16401, Rev 5.4, pages 12-14References provided: None Learning Objective: LO-PP-16401-02 Descr ibe the function of RCP seals 1, 2, and 3 including DP across each seal and
expected flow rate.
LO-PP-16401-03 Describe the co ntrol room indi cations for a failure of a RCP seal.LO-TA-60015 Respond to a Pa rtial Loss of Flow per 18005-CLO-TA-16009 Respond to abnormal RCP seal per 13003-1/2 Question origin:
MODIFIED - HL14 NRC Question #015/017G2.4.4Cognitive Level: C/A 10 CFR Part 55 Content: 43.5
Comments:
- JAT 12/19/13 (U/E)Early submittal 401-9 response:
Amanda's comments incorporated to include re moving the VCT pressure and trend in the stem and moved the question to procedure know ledge as opposed to Tech Spec bases.- JAT 2/4/14 (U/E)Response following revision from early submittal:
The first part of the question is improved. However, I am having difficulty seeing the TS completion time connection to the second part of the quest ion, and the way it's written, it appears as though the ques tion can be answered solely by knowing what specific dire ction is contained within the ARP (which is li kely not SRO-only , because it does not involve selection of procedures or sections of a procedure, nor does it require knowledge of >1h TS.). - JCC 2/5/14 Question replaced with modi fied question from HL14 NRC associated with RCP seal abnormality and procedureYou have completed the test!Tuesday, February 25, 2014 8:43:22 AM 4
- 1. 015/017G2.4.4 002/1/1/RCP MALF - EOP ENTRY/C/A - 4.3/MODIFIED/SRO/HL-14 NRC/TNT / RLM Unit 2 is at 12% power when the following annunciators are received.
- ALB08B05 "RCP # 3 CONTROLLED LKG HI / LO FLOW" - ALB08B04 "RCP # 3 NO. 2 SEAL LKOF HI FLOW" The OATC reports the following indications:
- RCP # 3 seal leakoff flow Hi Range meter is 6.0 gpm.
- RCP # 3 seal injection flow is 9.9 gpm. - RCP # 3 Seal Water Inlet temperature is 223
°F and stable.
Which one of the the following is the correct procedurally directed action(s) for the SS to take?Per 12004-C, "Power Operations (Mode 1)", comm ence a unit shutdown to be in Mode 3 in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Trip the reactor and enter 19000-C, "E-0 Reacto r Trip or Safety Injection", per 13003-2, "RCP Operation", stop RCP # 3 and close seal leakoff valve HV-8141C.
Per 13003-2, stop RCP # 3, close seal leakoff valve HV
-8141C, enter 18005-C,"Partial Loss of RCS Flow", commence unit shut down per 12004-C.
Per 12004-C, maintain reactor power at 25%, monitor the RCP per 13003-2 section 4.2.1 "Pump Operation With A Seal Abnormality", contact Duty Engineering.
A.B.C.D.Wednesday, February 05, 2014 12:44:26 PM 1
Approved By Vogtle Electric Generating Plant Procedure Version J.B. Stanley 18005-C 11.1 Effective Date PARTIAL LOSS OF FLOW Page Number 08/15/2012 1 of 5 Printed February 5, 2014 at 10:22 ABNORMAL OPERATING PROCEDURE CONTINUOUS USE PURPOSE This procedure addresses the lo ss of forced RCS flow during po wer operation below the P-8 (48%) setpoint.
SYMPTOMS ALB11-E04 RCP TRIP ALB12-A01(B01, C01, D01) RCP LOOP 1 (2, 3, 4) LOW FLOW ALERT ALB08-A01(B01, C01, D01) RCP 1 (2, 3, 4) MTR OVERLOAD ALB11-E06 UNDERVOLTAGE RCP BUS ALERT ALB11-F06 UNDERFREQUENCY RCP BUS ALERT UNIT 1 ALB33-A01(A02) 13.8KV SWGR 1NAA(1NAB) TROUBLE UNIT 2 ALB33-A01(A02) 13.8KV SWGR 2NAA(2NAB) TROUBLE
MAJOR ACTIONS Stabilize plant conditions. Shutdown to Mode 3. Restart RCP. Select appropriate UOP.
__
Approved By Vogtle Electric Generating Plant Procedure Version J.B. Stanley 18005-C 11.1 Effective Date PARTIAL LOSS OF FLOW Page Number 08/15/2012 3 of 5 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED Printed February 5, 2014 at 10:22
- 1. Check Reactor power - LESS THAN OR EQUAL TO 15%.
- 1. Perform the following:
1 a. Trip the Reactor.
1.a b. Go to 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION.
1.b 2. Stop any power changes in progress.
- 2. 2 3. Initiate the Continuous Actions Page.
- 3. 3 *4. Check affected loop SG NR Level - TRENDING TO 65%.
- 4. Control feed flow to maintain affected loop SG NR level between 60% and 70%.
4 5. Check Tavg - TRENDING TO PROGRAM. 5. Adjust control rods to restore Tavg.
5 6. Verify PRZR level - TRENDING TO PROGRAM. 6. 6 7. Verify PRZR pressure - TRENDING TO 2235 PSIG.
RUNNING. 8. Close the affected loop spray valve:
8 Loop 1: PIC-0455C Loop 4: PIC-0455B
- 9. Initiate shutdown to Mode 3 by initiating 12004-C, POWER OPERATION (MODE 1). (TS 3.4.4)
- 9. 9 10. Determine and correct the cause of the pump trip.
- 10. 10 11. Check shutdown to Mode 3 - COMPLETE.
- 11. Return to Step 9.
11 S Approved By M.G. Brill Vogtle Electric Generating Plant Procedure Version 13003-1 47.1 Effective Date 06/12/2013 REACTOR COOLANT PUMP OPERATION Page Number 13 of 42 INITIALS Printed February 5, 2014 at 10:24
4.2 SYSTEM
OPERATION 4.2.1 Pump Operation With A Seal Abnormality 4.2.1.1 IF the Plant Co mputer is available, trend the computer data points listed in Table 2. ________
4.2.1.2 IF the Plant Computer is NOT available, perform the following
- a. Monitor the QMCB indication listed in Table 2 at least hourly for the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. ________
- b. IF NO further seal degradation exists after 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, consult with the Shift Supervisor (SS) for less frequent monitoring. ________
4.2.1.3 Monitor the No. 1 seal for further degradation using Figure 1 and RCP Trip Criteria as follows:
- a. Evaluate the monitored indications using Figure 1, "RCP Seal Abnormalities Tree." ________
- b. IF evaluation of the monitored indications using Figure 1 requires immediate pump shutdown, Go to Step 4.2.1.4. ________
- c. IF any of the following RCP Trip Criteria is exceeded, Go To Step 4.2.1.4 for immediate RCP shutdown. ________
RCP TRIP CRITERIAMotor bearing temperature >195°F Motor stator-winding temperature >311°F Seal water inlet temperature >230°F RCP shaft vibration 20 mils RCP Frame vibration 5 mils #1 seal Differential Pressure <200 psid
- 1 seal leakoff flow (sum of #1 seal leakoff as indicated on the MCB and #2
seal leakoff read locally in containment)
< minimum on Figure 2 with pump bearing / seal inlet temperature
increasing Total loss of ACCW for a duration of 10 minutes Approved By M.G. Brill Vogtle Electric Generating Plant Procedure Version 13003-1 47.1 Effective Date 06/12/2013 REACTOR COOLANT PUMP OPERATION Page Number 14 of 42 INITIALS Printed February 5, 2014 at 10:24 d. WHEN directed by Figure 1, stop the affected RCP within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> as follows:
(1) Establish 9 gpm or greater seal injection flow to the affected pump. ________
(2) Stop the affected RCP by continuing with Step 4.2.1.4. ________
4.2.1.4 WHEN directed by the SS, perform an RCP shutdown as follows:
- b. IF Reactor Power is greater than 15% Rated Thermal Power: (1) Trip the Reactor and initiate 19000-C, "E-0 Reactor Trip Or Safety Injection." ________
(2) WHEN the immediate operator actions of 19000-C are complete, Go to Step 4.2.1.4.d. ________
- c. IF Reactor Power is less than 15% Rated Thermal Power, initiate 18005-C, "Partial Loss Of Flow." ________
- d. Stop the RCP by placing the RCP Non-1E Control Switch in STOP and then placing the RCP 1E Control Switch in
STOP: RCP Non-1E Control Switch 1E Control Switch Loop 1 1-HS-0495B 1-HS-0495A ________
Loop 2 1-HS-0496B 1-HS-0496A ________
Loop 3 1-HS-0497B 1-HS-0497A ________
Loop 4 1-HS-0498B 1-HS-0498A ________
Approved By M.G. Brill Vogtle Electric Generating Plant Procedure Version 13003-1 47.1 Effective Date 06/12/2013 REACTOR COOLANT PUMP OPERATION Page Number 15 of 42 INITIALS Printed February 5, 2014 at 10:24 CAUTION IF RCP #1 or #4 is stopped, the associated Spray Valve is placed in manual and closed to prevent spray short cycling. e. IF RCP #1 OR #4 is stopped, verify its associated spray valve is placed in MANUAL and closed.
RCP 1: 1-PIC-0455C ________
RCP 4: 1-PIC-0455B ________
- f. WHEN the RCP comes to a complete stop (as indicated by reverse flow), close the RCP Seal Leakoff Isolation valve for the affected pump.
RCP 1: 1-HV-8141A ________
RCP 2: 1-HV-8141B ________
RCP 3: 1-HV-8141C ________
RCP 4: 1-HV-8141D ________
- g. Secure the associated RCP Oil Lift Pump. ________
- h. IF RCP shutdown was due to loss of RCP seal cooling, review Limitation 2.2.11 for recovery action. ________
Approved By M.G. Brill Vogtle Electric Generating Plant Procedure Version 13003-1 47.1 Effective Date 06/12/2013 REACTOR COOLANT PUMP OPERATION Page Number 38 of 42 Printed February 5, 2014 at 10:24 FIGURE 1 - RCP SEAL ABNORMALITIES DECISION TREE Note 1: Abnormal Operating Range of Figure 2 Note 2: Non-operating Range of Figure 2 Note 3: ALB08 A-04, B-04, C-04 or D-04
Approved By M.G. Brill Vogtle Electric Generating Plant Procedure Version 13003-1 47.1 Effective Date 06/12/2013 REACTOR COOLANT PUMP OPERATION Page Number 39 of 42 Printed February 5, 2014 at 10:24 FIGURE 2
- 1. If the No. 1 seal leak rates are outside the normal (1.0-5.0 gpm) but within the operating limits ((0.8-5.5 gpm), continue pump operation. VERIFY that seal
injection flow exceeds No. 1 seal leak rate for the affected RCP. Closely monitor
pump and seal parameters and contact engineering for further instructions.
IF the No.1 seal leak off is between 0.6 and 0.8 gpm within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> determine the leak off with the No.1 plus No. 2 seals. IF the total leakoff is less than 0.8 gpm perform an orderly shutdown of the pump (see Note 4) within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
At 0.6 gpm on the No. 1 seal immediately shutdown the pump. (The 0.8 gpm and 0.6 gpm value includes the #2 seal leak off value from containment as well.)
Approved By M.G. Brill Vogtle Electric Generating Plant Procedure Version 13003-1 47.1 Effective Date 06/12/2013 REACTOR COOLANT PUMP OPERATION Page Number 40 of 42 Printed February 5, 2014 at 10:24
- 2. Minimum startup requirements are 0.2 gpm at 200 PSID differential across the No. 1 seal. For startups at differential pressures greater than 200 PSID, the
minimum No. 1 seal leak rate requirements are defined in the NO. 1 SEAL
NORMAL OPERATING RANGE (e.g., at 1000 psi differential pressure, do not
start the RCP with less than 0.5 gpm).
- 4. Per Westinghouse Technical Bulletin ESBU-TB-93-01-R1, total #1 seal leakoff is the sum of #1 seal leakoff and #2 seal leakoff. #1 seal leakoff is read directly at
the MCB and #2 seal leakoff can be obtained from instrumentation in
Containment.
Approved By J.B. Stanely Vogtle Electric Generating Plant Procedure Number Rev 17008-1 18 Date Approved 07/08/11 ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 08 ON PANEL 1A2 ON MCB Page Number 3 of 55 Printed February 5, 2014 at 10:18
(1) (2) (3) (4) (5) (6)
RCP 1 STANDPIPE
HI LEVEL RCP 1 NO. 2 SEAL LKOF HI FLOW RCP 1 CONTROLLED LKG HI/LO FLOW
RCP 2 STANDPIPE
HI LEVEL RCP 2 NO. 2 SEAL LKOF HI FLOW RCP 2 CONTROLLED LKG HI/LO FLOW
RCP 3 STANDPIPE
HI LEVEL RCP 3 NO. 2 SEAL LKOF HI FLOW RCP 3 CONTROLLED LKG HI/LO FLOW
RCP 4 STANDPIPE
HI LEVEL RCP 4 NO. 2 SEAL LKOF HI FLOW RCP 4 CONTROLLED LKG HI/LO FLOW
RCP 1 VIBRATION
ALERT
RCP 2 VIBRATION
ALERT RCP VIBRATION
HIGH
RCP SEAL WATER INJ FILTER HI P F
RCP 3 VIBRATION
ALERT
RCP 4 VIBRATION
ALERT RCP VIB MON PNL TROUBLE RCP SEAL WATER INJ
LO FLOW Approved By J.B. Stanely Vogtle Electric Generating Plant Procedure Number Rev 17008-1 18 Date Approved 07/08/11 ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 08 ON PANEL 1A2 ON MCB Page Number 10 of 55 Printed February 5, 2014 at 10:18 WINDOW A04
ORIGIN SETPOINT 1-FIS-0194
0.9 gpm
RCP 1 NO. 2 SEAL LKOF
HI FLOW
1.0 PROBABLE
CAUSE
- 1. Number 2 Seal failure.
- 2. Sudden reduction in RCDT level or pressure.
2.0 AUTOMATIC
ACTIONS NONE
3.0 INITIAL
OPERATOR ACTIONS
- 1. Check RCDT pressure on 1-PISL-9699 (QPCP) 3 psig or greater.
- 2. Dispatch Operator to check RCDT pressure and level at PLPP:
- a. Pressure 2-3 psig,
- b. Level 20-75%.
- 3. IF RCDT pressure and level are normal, Go To 13003-1, "Reactor Coolant Pump Operation" for instructions on RCP operation with seal malfunctions.
4.0 SUBSEQUENT
OPERATOR ACTIONS NONE
5.0 COMPENSATORY
OPERATOR ACTIONS NONE
END OF SUB-PROCEDURE
REFERENCES:
1X4DB114, 1X6AB09-119, PLS Approved By J.B. Stanely Vogtle Electric Generating Plant Procedure Number Rev 17008-1 18 Date Approved 07/08/11 ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 08 ON PANEL 1A2 ON MCB Page Number 11 of 55 Printed February 5, 2014 at 10:18 WINDOW A05
ORIGIN SETPOINT 1-FT-0161 1-FT-0157
4.8 gpm 0.8 gpm
RCP 1 CONTROLLED LKG
HI/LO FLOW
1.0 PROBABLE
CAUSE
- 1. High Flow:
- a. Flashing in the Seal Leakoff Line due to loss of seal injection flow or high seal injection temperature,
- b. Failure of Number 1 Seal.
- 2. Low Flow:
- a. Low differential pressure across Number 1 Seal,
- b. High Volume Control Tank (VCT) pressure,
- c. Excess letdown in service,
- d. Failure of Number 2 Seal.
2.0 AUTOMATIC
ACTIONS NONE
Approved By J.B. Stanely Vogtle Electric Generating Plant Procedure Number Rev 17008-1 18 Date Approved 07/08/11 ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 08 ON PANEL 1A2 ON MCB Page Number 12 of 55 Printed February 5, 2014 at 10:18 WINDOW A05 (Continued)
3.0 INITIAL
OPERATOR ACTIONS NOTE RCP 1 No. 1 seal water leakoff high range flow may be monitored using computer point F0161.
- 1. Observe seal injection flow and seal leakoff flow, as well as excess letdown temperature and pressure for indication of an actual seal
anomaly. 2. IF a seal problem is indicated, Go To 13003-1, "Reactor Coolant Pump Operation".
- 3. IF an instrument problem is indicated, initiate maintenance as required.
4.0 SUBSEQUENT
OPERATOR ACTIONS NONE
5.0 COMPENSATORY
OPERATOR ACTIONS
- 1. Verify proper seal leakoff using 1-FI-0156A and 1-FI-0160A once per shift, and refer to 13003-1, "Reactor Coolant Pump Operation" if leakoff is outside the limits.
- 2. Log corrective actions to repair the disabled annunciator or reasons for no action on 10018-C, "Annunciator Control", Figure 2.
- 3. Log compensatory actions on 10018-C, "Annunciator Control", Figure 5.
END OF SUB-PROCEDURE
REFERENCES:
1X4DB114, 1X6AB09-119, PLS V-LO-PP-16401 12Objective 4Demonstrate using the seal package model.The seal package consist of three seals 1)Number 1 seal(film riding)a) The primary seal b) Seal is accomplished with a hydrostatic film between the shaft runner and seal ring.c) No mechanical contact between seal ring and shaft runner (must keep ~P >200 psid)
- 2) Number 2 seal (face rubbing)a) Provides back up for #1 seal b) Consist of carbon graphite (face rubbing seal)c) Graphite makes contact with runner which rotates with shaft d) If #1 seal fails , #2 seal converts to a film riding seal if #1 seal leak off valve is closed and seal is exposed to full RCS pressure. #2 seal designed to allow plant shutdown and should last approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.e) Placing #2 seal in service with the RCP shaft still rotating will tend to score the shaft at the #2 seal area. This can require extensive repairs before placing the RCP back in service. Vogtle chooses toremove RCP from service and allow its shaft to come to a standstill before closing the #1 seal leakoff valve to avoid this problem.
- 3) Number 3 seal (face rubbing)a) Prevents the leakage of liquid and gases from the RCS into containment.b) Consist of carbon graphite seal which makes contact with runner (face rubbing)c) The runner is around the shaft and rotates with it.
d) The seal is actually two graphite sealing surfaces called dams.
Objective 1d2) RCP Motor AuxiliariesA) Motor Cooler 1) Containment Air is drawn into the motor by fan blades on motor's rotor
- 2) It is then exhausted through the motor cooler
- 3) ACCW is the cooling medium used in the cooler (cools the outgoing air)4) This arrangement limits containment air temperature rise and in turn limits motor temperature.
- 3) FlywheelA) Addressed in tech spec administrative sectionB) Stores rotational energy of the pump and motor while running then releases energy by maintaining pump motion to slowly reduce core flow following loss of power for core protection.4) RCP motor space heaterA) Each RCP motor has a electric resistance heater.
B) Used to prevent moisture accumulation in windings when motor is shutdown.
C) Not needed when motor is in operation because of heat generation from motor windings.
D) Heaters are automatically energized when either of the RCP motor breakers are opened.
E) Heaters are supplied from 480 V MCCsRCP motor burned up at Vogtlethat was attributed to moisture from space heater breaker being open when pump was shutdown. The space heater did not energize therefore moisture accumulated while the pump was shutdown during the outage. Upon restart the motor windings shorted out. Motor rebuild was required.V-LO-PP-16401 13 V-LO-PP-16401 14Objective 2RCP Seal Injectiona)Provided from CVCS b)8 gpm per pump c)5 gpm is directed through to lower pump radial bearing and into the RCS loop.d)The remaining 3 gpm supplies #1 and #2 seals e)#3 seal injection is from small tanks called Standpipes. (Gravity Fed)f)Flow path1) 8 gpm from CVCS enters RCP at 2250 psig
- 2) 5 gpm passes through the lower pump bearing lubricating and cooling it.
- 3) Seal injection at 2250 psig prevents RCS water from escaping the loop.4) 3 gpm is directed through #1 seal
- 5) A pressure drop at 2220 psid across the #1 seal occurs.
- 6) Approximately 3 gph (0.05 gpm) leak off from #1 seal is used as seal injection to #2 seal.7) The remainder of #1 seal leak off is directed back to the VCT via seal water return.8) 3 gph passes through #2 seal and the leak off is directed to the RCDT (~5-6 psig)9) 800 cc/hr seal injection for #3 seal is provided by standpipe (~10 psig)10) The standpipes are located at a higher elevation than the RCP and gravity feeds #3 seal; standpipes Auto fill from RMWST.11) #3 seal injection is injected between the two dams and sealing surfaces.12) #3 seal injection pressure is slightly higher than #2 seal injection leak off.13) This prevents RCS liquids or gases from escaping to the containment environment.14) #3 seal has two leak off paths a) The outer dam leak off (400 cc/hr) combines with #2 seal leak off and is routed to RCDTb) The inner dam leak off (400 cc/hr) is directed to the containment sump. SMART -Solid Knowledge. If the #1 seal leakoff was isolated, the #2 seal would become a film riding seal due to increased pressure across the #2 seal facing.
- 1. 008A2.02 001/LOCT AND LOIT/SRO/C/A 3.2/3.5/008A2.02/LO-TA-60026//HL-18 NRC/
Initial conditions:
- Unit 1 is at 100% reactor power.
- CCW pump #5 is tagged out for maintenance.
- 1-LSLL-1854, CCW Surge Tank level switch for CCW pump #3, has failed. - ALB02-A05 CCW TRAIN A SURGE TK LO-LO LVL is received.
Current conditions:
- ALB02-A06 CCW TRAIN A LO HDR PRESS is received. - ALB02-B06 CCW TRAIN A LO FLOW is received.
Which one of the following completes the following statement?
Demin Water Makeup Valve to the CCW Train 'A' Surge Tank __(1)__ automatically open,and per Tech Spec 3.7.7, "Com ponent Cooling Wate r (CCW) System," Train 'A' CCW is declared __(2)__. __(1)__ __(2)__ will OPERABLE will inoperable will NOT OPERABLE will NOT inoperable A.B.C.D.008 Component Cooling Water System (CCWS)A2.02 Ability to (a) predict the impacts of the following malfunctions oroperations on the CCWS and (b) based on those pr edictions, useprocedures to correct, control, or mitigate the consequences ofthose malfunctions or operations:- High / Low surge tank levelK/A MATCH ANALYSIS The candidate is presented with a plausible scenario where a CC W Surge Tank levelThursday, February 20, 2014 3:54:06 PM 1
transmitter fails low and is required to determine the impact of the failure on the system.
With one CCW pump tagged out and the inability to start the pump with the failed level transmitter, a decision of operability must be deter mined which is SRO requiredknowledge. EXPLANATION OF REQUIRED KNOWLEDGE Each train of CCW is comprised of 3 pu mps. Two pumps are required to met LCO 3.7.7. The pumps automatically start on SI, LOSP, Lo header pre ssure, and tr ip of a running pump. Only the SI and LOSP starts are required per SR 3.7.7.2.
When level switch 1-LSLL-1854 fails low, CCW pum p #3 trips. With CCW pump #5already tagged out, the system header pressure lowers and annunciators ALB02-A06 &
B06 alarm. Since automati c Demin Water Makeup Valve LV
-1850 is controlled by level switches 1-LSL-1850 & 1-LSH-1850, its operation is not affected. (Reference P&ID 1X4DB136) No actual low level condition exists, theref ore LV-1850 will not open.
Since 1-LSLL-1854 failing low resulted in a pump trip, SR 3.7.7.2 to verify each CCW pump starts automatically on an actual or simulated actuation signal can not be satsified. Per TS SR 3.0.1, failure to meet a Survelliance, wh ether such failure is experienced during the performance of the Survelliance or between performances ofthe Survelliance, shall be failure to meet the LCO. Therefore, CCW Pump #3 must be declared inoperable. Since CCW Pump #5 was already inoperable , TS 3.7.7 LCO isnot met and RAS 'A' must be entered.ANSWER / DISTRACTOR ANALYSISA. Incorrect. Plausible. The first part is incorrect.
Since automatic makeup valve LV-1850 is controlled by level switches 1-LSL-1850 &
1-LSH-1850, its oper ation is not affected. No actual low level condition exists, therefore LV-1 850 will not open. However, if the candidate believes makeup is controlled off th e same leveltransmitter or misses that t he level transmitter failed and believes there is an actual lo level, then makeup valve LV-1850 would be expected to open.
The second part is incorrect.
SR 3.7.7.2 can not be met andCCW Pmp #3 and #5 are both in operable. LCO 3.7.7 cannot be met. However, a candi date not familiar with the requirements of SR 3.7.72 & SR 3.0.1 or the pump start logic could conclude that the non-safety related surge tank low level trip would be bypassed on an SI or LOSP or that auto makeup would restore a low level co ndition and CCW Pmp #3 would remain OPERABLE.
B. Incorrect. Plausible. Th e first part is incorre ct. See the first part of choice A above.
The second part is correct. SR 3.7.7.2 can not be met andCCW Pmp #3 and #5 are both in operable. LCO 3.7.7 cannot be met.Thursday, February 20, 2014 3:54:06 PM 2
C. Incorrect. Plausible. The first part is correct. Since automatic makeup valve LV-1850 is controlled by level switches 1-LSL-1850 &
1-LSH-1850, its oper ation is not affected. No actual low level condition exists, therefor e LV-1850 will not open.
The second part is inco rrect. See the second part of choice A above.D. Correct. The first part is correct.
See the first part of choice C above.
The second part is correct. See the second part of choice B above.SRO JUSTIFICATION (10CFR43(b))(2) Facility operating limitations in the technical specifications and their bases.-Can question be answered solely by knowing = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action?
No,generic LCO SR3.0.1 knowledge is required. No 1 hr or less actions exist.-Can question be answered solely by knowing the LCO/TRM information listed "above-the-line?"
No, the required knowledge is found in generic LCO SR3.0.1and SR 3.7.7.2, which is below the line.
-Can question be answered solely by knowing the TS Safety Limits?
No, this question does not involve a TS Safety Limit.-Does the question involve one or more of the following for TS,TRM, or ODCM?
- Application of Required Actions (Sec tion 3) and Surve illance Requirements (Section 4) in accordance with rules of applicati on requirements (Section 1)* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR4.0.1 thru 4.0.4)
Yes, the question requires an OPERABILITY call be madeby utilizing generic LCO SR3.0.1 as applied to SR 3.7.7.2.* Knowledge of TS bases that is required to analyze TS required actions and terminologyThursday, February 20, 2014 3:54:07 PM 3
Level: SROTier # / Group # T2 / G1 K/A# 008A2.02 Importance Rating: 3.2 / 3.5 Technical
Reference:
ARP 17002-1, Rev 24.1, pages 3, 14, & 26 P&ID 1X4DB136, Rev 33.0
TS 3.7.7, Amendment No. 96, pages 3.7.7-1 & 2 TS 3.0.1, Amendment No. 125, page 3.0-4References provided: None
Learning Objective:
LO-PP-10101-04 From memory, describe the expectedsystem response and operator corrective
actions for each of the following:
- d. Surge Tank Low Level LO-LP-39211-04 Descr ibe the bases for any given Tech Spec in section 3.7.LO-TA-60026 Respond to a Loss of CCW per 18020-C LO-TA-10006 Draw the CCW SystemQuestion origin: BANK - Reus e - HL18 Question # 008A2.02Cognitive Level: C/A 10 CFR Part 55 Content: 43.2
Comments:
You have completed the test!Thursday, February 20, 2014 3:54:07 PM 4
CCW System
3.7.7 Vogtle
Units 1 and 2 3.7.7-1 Amendment No. 96 (Unit 1) Amendment No. 74 (Unit 2) 3.7 PLANT SYSTEMS
3.7.7 Component
Cooling Water (CCW) System
LCO 3.7.7 Two CCW trains shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CCW train inoperable.
A.1 -------------NOTE-------------
Enter applicable
Conditions and Required
Actions of LCO 3.4.6, "RCS Loops - MODE 4,"
loops made inoperable by CCW.
Restore CCW train to OPERABLE status.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B. Required Action and associated Completion
Time of Condition A not met. B.1 Be in MODE 3.
AND B.2 Be in MODE 5.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />
CCW System
3.7.7 Vogtle
Units 1 and 2 3.7.7-2 Amendment No. 158 (Unit 1) Amendment No. 140 (Unit 2) SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.7.1 ------------------------------NOTE---------------------------- Isolation of CCW flow to individual components does not render the CCW System inoperable.
Verify each CCW manual, power operated, and
automatic valve in the flow path servicing safety
related equipment, that is not locked, sealed, or otherwise secured in position, is in the correct position.
In accordance with the Surveillance Frequency Control Program SR 3.7.7.2 Verify each CCW pump starts automatically on an actual or simulated actuation signal.
In accordance with the Surveillance Frequency Control Program
SR Applicability
3.0 Vogtle
Units 1 and 2 3.0-4 Amendment No. 125 (Unit 1)
Amendment No. 103 (Unit 2) 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3. Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.
SR 3.0.2 The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.
For Frequencies specified as "once," the above interval extension does not apply.
If a Completion Time requires periodic performance on a "once per . . ."
basis, the above Frequency extension applies to each performance after the initial performance.
Exceptions to this Specification are stated in the individual Specifications.
SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the risk impact shall be managed.
If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.
(continued)
Approved By P.H. Burwinkel Vogtle Electric Generating Plant Procedure Version 17002-1 24.1 Effective Date 07/27/2012 ANNUNCIATOR RESPONSE PROCE DURES FOR ALB 02 ON PANEL 1A1 ON MCB Page Number 3 of 42 Printed February 20, 2014 at 15:04 (1) (2) (3) (4) (5) (6)
A NSCW TRAIN A
F-1 HI VIB
NSCW TRAIN A
F-2 HI VIB
NSCW TRAIN A
F-3 HI VIB
NSCW TRAIN A
F-4 HI VIB
CCW TRAIN A
SURGE TK LO-LO LVL
CCW TRAIN A
B NSCW TRAIN A
NSCW TRAIN A TRANSF PMP LO DISCH PRESS
CCW TRAIN A
SURGE TK HI/LO LVL
C NSCW TRAIN A CLG TWR BASIN
HI/LO LVL
NSCW TRAIN A
NSCW TRAIN A
CCW TRAIN A
SURGE TK MAKE UP LVL
CCW TRAIN A
NSCW TRAIN A
NSCW INTERTIE TRN A TO TRN B HI FLOW
CCW TRAIN A
NSCW TRAIN A
NSCW TRAIN A NORM/BYP VLV MISPOSITIONED
RMWST VAC DEGASIFIER PNL ALARM
LO FLOW PRIMARY EQUIPMENT
HI TEMP F
NSCW TRN A RX
CVTY CLG COIL LOW FLOW
RX MAKE UP STOR TK LO-LO LVL
RX MAKE UP STOR TK HI/LO LVL
Approved By P.H. Burwinkel Vogtle Electric Generating Plant Procedure Version 17002-1 24.1 Effective Date 07/27/2012 ANNUNCIATOR RESPONSE PROCE DURES FOR ALB 02 ON PANEL 1A1 ON MCB Page Number 14 of 42 Printed February 20, 2014 at 15:04 WINDOW A05 ORIGIN SETPOINT 1-LSLL-1852 1-LSLL-1854
1-LSLL-1856
4.75 in. below CL (equal to 42%)
CCW TRAIN A SURGE TK LO-LO LVL
1.0 PROBABLE
CAUSE
- 1. Failure of automatic make-up from Demineralized Water System.
- 2. Failure of manual make-up from Reactor Makeup Water System.
- 3. Leak in Component Cooling Water System.
2.0 AUTOMATIC
ACTIONS LO-LO level will trip Component Cooling Water Pumps.
3.0 INITIAL
OPERATOR ACTIONS Go To 18020-1, "Loss Of Component Cooling Water."
4.0 SUBSEQUENT
OPERATOR ACTIONS NONE
5.0 COMPENSATORY
OPERATOR ACTIONS NONE
END OF SUB-PROCEDURE
REFERENCES:
1X4DB136, 1X3D-B D-L01A, 1X3D-BD-L01C, 1X3D-BD-L01E, 1X5DN091-1, -2, -3, 1X5DT0022, CX5DT101-96 Approved By P.H. Burwinkel Vogtle Electric Generating Plant Procedure Version 17002-1 24.1 Effective Date 07/27/2012 ANNUNCIATOR RESPONSE PROCE DURES FOR ALB 02 ON PANEL 1A1 ON MCB Page Number 26 of 42 Printed February 20, 2014 at 15:04 WINDOW C05 ORIGIN SETPOINT 1-LSL-1850
1.25 in. above CL (equal to 52%)
CCW TRAIN A SURGE TK MAKE UP LVL
1.0 PROBABLE
CAUSE Component Cooling Water (CCW) System leakage.
2.0 AUTOMATIC
ACTIONS Makeup Valve 1-LV-1850 opens.
3.0 INITIAL
OPERATOR ACTIONS NONE
4.0 SUBSEQUENT
OPERATOR ACTIONS
- 1. Monitor level using 1-LIT-1846 or computer point L2671.
- 2. IF 1-LV-1850 fails to open:
- a. Open the valve using 1-HS-1850 on QMCB, b. Continue to monitor level, c. Open 1-LV-1848 using 1-HS-1848 if level continues to fall.
- 3. IF equipment failure is indicated, initiate maintenance as required.
5.0 COMPENSATORY
OPERATOR ACTIONS NONE
END OF SUB-PROCEDURE
REFERENCES:
1X4DB136, 1X3D-BD-L01G, 1X5DT0022, CX5DT101-95
- 1. 008AG2.4.41 001/LOCT AND LOIT/SRO/C/A 2.9/4.6/008AG2.4.41/LO-TA-40002///
Initial condition: - Unit 1 is at 100% reactor power.
Current conditions:
- ALB12-F01 PRZR SAFETY RELIEF DISCH HI TEMP is received. - PRT level, temperatur e, and pressure are all increasing.
- CVCS char ging flow is 110 gpm.
- All RCP seal inje ction and #1 seal return flows are within normal operating range.
- CVCS letdown is isolated.
- Pressurize r level is 55% and stable.
Which one of the following completes the following statement?
Per Tech Spec 3.4.13, "RCS Operationa l Leakage," the RCS leakage is classified as
__(1)__, and per NMP-EP-110, "Emergency Cl assification Determination and Initial Action," the Shift Manager is required to declare as a minimu m a(n) __(2)__.REFERENCE PROVIDED __(1)__ __(2)__ identified NOUE identified Alert unidentified NOUE unidentified Alert A.B.C.D.K/A 008 Pressurizer Vapor Space AccidentG2.4.41 Knowledge of the emergency action level thresholds and classifications.K/A MATCH ANALYSIS The question sets up plausible scenario w here Main Cont rol annunciators are received and various plant parameters and indications are provided. Based on the information provided the SRO candidate is required to asse ss the type of leakage indicated and determine if any emer gency action level th resholds were ex ceeded. The type ofFriday, February 21, 2014 10:13:47 AM 1
leakage reflects the first part of the KA requirement for a vapor space accident, the second action to quantify the leakage amount and dete rmine any Emergency Plan requirements meets the sec ond part of the KA and brings the question to the SROknowledge level. EXPLANATION OF REQUIRED KNOWLEDGE The given plant conditions are indicitive of a Pr essuirzer Code Safety valve relieving into the PRT. Per TS 3.4.13 Bases, id entified leak is defined as being from a specifically known and located source, ie it must be collected and quantifiable, and not pressure boundary. Based on the given cond itions, the leakrate is approximately 98 gpm, [charging-(letdown + seal leakoff)]. Total seal leakoff is normally about 12 gpm.
Since the leakage is across a va vle seat, it is not pressure boundary. Therefore, the leakage described fully meets the definition of IDENTIFIED.
Per NMP-EP-110-GL03 Figure 1, a potential loss of the RCS barrier exists if RCS leak rate is non-isolatabe and >120 gpm. Since le akrate is calculated to be 98 gpm, this threshold has not been excee ded. Per NMP-EP-110-GL02 Figure 2, NOUE threshold SU5 is exceeded if identified leakage is > than 25 gpm. T herefore, this threshold has been exceeded.ANSWER / DISTRACTOR ANALYSISA. Correct. Part 1 is correct it requi red the candidate to determine the type of RCS leakage based on informat ion provided and knowledge of the Tech Spec defin ition of the various leakage categories.
From the information the cand idate should determine this is identified leakage.
LEAKAGE, such as that from pump seals orvalve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collectionsystems or a sump or collecting tank.
Part 2 required the candidate to take the information gathered in Part 1 and apply it to the various Emergency Action Level classification thresholds. From that a leak rate of 98 gpm is determined and limi ts for NOUE rec ognized as exceeded.B. Incorrect. Plausible. Part 1 is correct. See Part 1 of choice A above.
Part 2 is incorrect but 'plausible' because the candidate may determine that 110 gpm charging added to 32 gpm seal injection (which is alreay included in the charging flow indicator reading) going into RCS is 142 GPM leak. This minus the RCP seal return fl ow of 3.2 each gpm would be above the ALERT threshold of 120 gpm for the potential loss of the RCS barrier. (Reference P&IDs 1X4DB114 and 1X4DB116-1)C. Incorrect. Plausible. Part 1 is incorrect bu t 'plausible' in that the candidate must first identify the type of RCS leak age. As the PRT continues torecieve effluent from the Przr Code Safety, it will eventuallyFriday, February 21, 2014 10:13:47 AM 2
rupture. If the candidate assumes that once this occur the leakage is no longer being "collected" in the PRT and not think about the leakage now bei ng colleted in the containment sump, then it would be reasonable to interpret the indications asUNIDENTIFIED RCS leakage.
Part 2 is correct. See Part 2 of choice A above.
D. Incorrect. Plausible. Part 1 is in correct. See Part 1 of choice C above. Part 2 is incorrect. See Part 2 of choice B above.SRO JUSTIFICATION(5) Assessment of facility conditions and selection of appropriate proceduresduring normal, abnormal, and emergency situations.-Can the question be answered solely by knowing "systems knowledge", i.e.,how the system works, flowpath, logic, component location?
No, the answerrequires specific knowledge of emergency classification thresholds.-Can the question be answered solely by knowing immediate operator actions? No, IOAs are not addressed by this question.
-Can the question be answered solely by knowing entry conditions for AOPs orplant parameters that require direct entry to major EOPs?
No, the question does not address AOP or EOP entry conditions.-Can the question be answered solely by knowing the purpose, overall sequenceof events, or overall mitigative strategy of a procedure?
No, the answer requiresspecific knowledge of emergency classification thresholds.-Does the question re quire one or more of the following?* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or se ction of a procedure to mi tigate, recover, or with which to proceed* Knowledge of when to impl ement attachments and appendices, including how to coordinate these it ems with procedure steps* Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific s ubprocedures or emergency contingency procedures* Knowledge of administrative procedures that specify hierarchy,implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
Yes, the answer requires specificknowledge of emergency cl assification thresholds anddetermination of the sp ecific classification based on current plantconditions. This is an SRO ONLY job link associated with an SRO
ONLY objective. [L O-LP-40101-13 Given an emergency scenario,and the procedure, classify the emergency (SRO only).]Friday, February 21, 2014 10:13:47 AM 3
Level: SROTier # / Group # T1 / G1 K/A# 008G2.4.41 Importance Rating: 2.9 / 4.6 Technical
Reference:
NMP-EP-110-GL03, Rev 3.0, page 58 NMP-EP-110-GL03, Figure 1, Rev 3.0, page 121 NMP-EP-110-GL03, Figure 2, Rev 3.0, page 122 P&ID 1X4DB114, Rev 50.0
P&ID 1X4DB116-1, Rev 41.0 TS 3.4.13 Bases, Rev 2-9/06, page B.3.4.13-2TS 3.4.13 Bases, Rev 1-9/03, page B.3.4.13-3 References provided:
NMP-EP-110-G L03, Figure 1 NMP-EP-110-G L03, Figure 2 NMP-EP-110-G L03, Figure 3Learning Objective: LO-TA-40002 Emergency Classification andImplementing Instructions using
NMP-EP-110 (SRO Only)LO-TA-60014 Respond to Reactor Coolant System Leakage per 18004-C LO-LP-40101-13 Given an em ergency scenario, and the procedure, classify the emergency (SRO
only).LO-LP-39202-02 Demonstrate a working knowledge of the application of all Technical Specification
definitions.
LO-LP-39202-01 Define the fo llowing terms, as per Plant Vogtle Tech Specs:
- i. identified leakage
- q. unidentified leakage LO-LP-60304-04 Given the sym ptoms of RCS leakage intoan area or system, correctly identify the leakage area or system.
LO-LP-60304-10 Given condit ions and/or indications of leaks identified in A ttachment "A" of AOP 18004-C, determine t he probable location of the leakage per 18004-C.
LO-LP-60304-12 Discuss how approximate RCS leak rate is determined.Question origin: NEW Cognitive Level: C/A 10 CFR Part 55 Content: 43.5 Comments: You have completed the test!Friday, February 21, 2014 10:13:47 AM 4
RCS Operational LEAKAGE B 3.4.13 (continued)
Vogtle Units 1 and 2 B 3.4.13-2 Rev. 2-9/06 BASES (continued)
APPLICABLE Except for primary to secondary LEAKAGE, the safety analyses SAFETY ANALYSES do not address operational LEAKAGE. However, other operational LEAKAGE is related to the safety analyses for LOCA; the amount of leakage can affect the probability of such an event. The safety analyses for an event resulting in steam discharge to the atmosphere assumes that primary to secondary LEAKAGE from all steam generators (SGs) is one gallon per minute or increases to one gallon per minute as a result of accident induced conditions. The LCO requirement to limit primary to secondary LEAKAGE through any one SG to less than or equal to 150 gallons per day is significantly less than the conditions assumed in the safety analysis.
The RCS operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).
LCO RCS operational LEAKAGE shall be limited to:
- a. Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of an off-normal condition. LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.
- b. Unidentified LEAKAGE One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this LCO could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary.
- c. Identified LEAKAGE Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of unidentified LEAKAGE and is well within the capability of the RCS Makeup System. Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary
RCS Operational LEAKAGE B 3.4.13 (continued)
Vogtle Units 1 and 2 B 3.4.13-3 Rev. 1-9/06 BASES LCO c. Identified LEAKAGE (continued)
LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE). Violation of this LCO could result in continued degradation of a component or system. d. Primary to Secondary LEAKAGE through Any One SG The limit of 150 gallons per day per SG is based on the operational LEAKAGE performance criterion in NEI 97-06, Steam Generator Program Guidelines (Ref. 4). The Steam Generator Program operational LEAKAGE performance criterion in NEI 97-06 states, "The RCS operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day." The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.
APPLICABILITY In MODES 1, 2, 3, and 4, the potential for RCPB LEAKAGE is greatest when the RCS is pressurized.
In MODES 5 and 6, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE.
LCO 3.4.14, "RCS Pressure Isolation Valve (PIV) Leakage," measures leakage through each individual PIV and can impact this LCO. Of the two PIVs in series in each isolated line, leakage measured through one PIV does not result in RCS LEAKAGE when the other is leak tight. If both valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable identified LEAKAGE.
NMP-EP-110-GL03 - VEGP EALs - ICs, Threshold Values and Basis Version 3.0 Figure 1 121 VOGTLE ELECTRIC GENERATING PLANT Figure 1 - Fission Product Barrier Evaluation NMP-EP-110- GL03 Rev 3.0 General Emergency Site Area Emergency Alert Unusual Event FG1 Loss of ANY Two Barriers AND Loss or Potential Loss of Third Barrier FS1 Loss or Potential Loss of ANY Two BarriersFA1 ANY Loss or ANY Potential Loss of EITHER Fuel Clad OR RCSFU1 ANY Loss or ANY Potential Loss of ContainmentFuel Clad Barrier Loss Potential Loss 1. Critical Safety Function Status Core-Cooling RED
- 1. Critical Safety Function Status Core Cooling-ORANGE OR Heat Sink-RED
- 2. Primary Coolant Activity Level Indications of RCS Coolant Activity greater than 300 Ci/gm Dose Equivalent I-131 2. Primary Coolant Activity Level Not Applicable
- 3. Core Exit Thermocouple Readings Core Exit TCs greater than 1200 F 3. Core Exit Thermocouple Readings Core Exit TCs greater than 711 F 4. Reactor Vessel Water Level Not Applicable
- 4. Reactor Vessel Water Level RVLS LEVEL less than 63% 5. Containment Radiation Monitoring Containment Radiation Monitor RE-005 OR 006 greater than 6E+6 mR/hr 5. Containment Radiation Monitoring Not Applicable 6. Other Indications Not applicable
- 6. Other Indications Not applicable 7. Emergency Director Judgment Judgment by the ED that the Fuel Clad Barrier is lost. Consider conditions not addressed and inability to determine the status of the Fuel Clad Barrier 7. Emergency Director Judgment Judgment by the ED that the Fuel Clad Barrier is potentially lost. Consider conditions not addressed and inability to determine the status of the Fuel Clad Barrier. RCS Barrier Loss Potential Loss 1. Critical Safety Function Status Not Applicable
- 2. RCS Leak Rate RCS subcooling less than 24F {less than 38 F Adverse} due to an RCS leak greater than Charging / RHR capacity
- 2. RCS Leak Rate Non-isolable RCS leak (including SG tube Leakage) greater than 120 gpm 3. SG Tube Rupture SGTR resulting in an SI actuation 3. SG Tube Rupture Not Applicable 4. Containment Radiation Monitoring CTMT Rad Monitor RE-005 OR 006 greater than 2.0E+4 mR/hr 4. Containment Radiation Monitoring Not Applicable 5. Other Indications Not applicable
- 5. Other Indications Unexplained level rise in ANY of the following: Containment sump Reactor Coolant Drain Tank (RCDT) Waste Holdup Tank (WHT) 6. Emergency Director Judgment Judgment by the ED that the RCS Barrier is lost. Consider conditions not addressed and inability to determine the status of the RCS Barrier 6. Emergency Director Judgment Judgment by the ED that the RCS Barrier is potentially lost. Consider conditions not addressed and inability to determine the status of the RCS Barrier. Containment Barrier Loss Potential Loss 1. Critical Safety Function Status Not Applicable
- 1. Critical Safety Function Status Containment-RED
- 2. Containment Pressure Rapid unexplained CTMT pressure lowering following initial pressure rise OR Intersystem LOCA indicated by CTMT pressure or sump level response not consistent with a loss of primary or secondary coolant 2. Containment Pressure CTMT pressure greater than 52 psig OR CTMT hydrogen concentration greater than 6%
OR CTMT pressure greater than 21.5 psig AND Less than the following minimum operable equipment: Four CTMT fan coolers AND One train of CTMT spray3. Core Exit Thermocouple Reading Not applicable
- 3. Core Exit Thermocouple Reading CORE COOLING CSF - RED OR - ORANGE for greater than 15min AND RVLS LEVEL less than 63%
- 4. SG Secondary Side Release with Primary to Secondary Leakage RUPTURED S/G is also FAULTED outside of containment OR Primary-to-Secondary leakrate greater than 10 gpm with nonisolable steam release from affected S/G to the environment 4. SG Secondary Side Release with P-to-S Leakage Not applicable 5. CNMT Isolation Valves Status After CNMT Isolation CTMT isolation valve(s)
OR damper(s) are NOT closed resulting in a direct pathway to the environment after containment isolation is required 5. CNMT Isolation Valves Status After CNMT Isolation Not Applicable 6. Significant Radioactive Inventory in Containment Not Applicable
- 6. Significant Radioactive Inventory in Containment CTMT Rad monitor RE-005 OR 006 greater than 2.4E+8 mR/hr 7. Other Indications Pathway to the environment exists based on VALID RE-2562C Alarm AND RE-12444C OR RE-12442C Alarms 7. Other Indications Not applicable
- 8. Emergency Director Judgment Judgment by the ED that the CTMT Barrier is lost. Consider conditions not addressed and inability to determine the status of the CTMT Barrier
- 8. Emergency Director Judgment Judgment by the ED that the CTMT Barrier is potentially lost. Consider conditions not addressed and inability to determine the status of the CTMT Barrier
NMP-EP-110-GL03 - VEGP EALs - ICs, Threshold Values and Basis Version 3.0 122 FIGURE 2 NMP-EP-110-GL03 - VEGP EALs - ICs, Threshold Values and Basis Version 3.0 58 SU5 Initiating Condition RCS Leakage.
Operating Mode Applicability: Power Operation (Mode 1) Startup (Mode 2)
Hot Standby (Mode 3) Hot Shutdown (Mode 4)
Threshold Values: (1 OR 2) 1. Unidentified OR pressure boundary leakage greater than 10 gpm.
- 2. Identified leakage greater than 25 gpm.
Basis: This IC is included as a NOUE because it may be a precursor of more serious conditions and, as result, is considered to be a potential degradation of the level of safety of the plant. The 10 gpm value for the unidentified and pressure boundary leakage was selected as it is observable with normal control room indications. Lesser values must generally be determined through time-consuming surveillance test s (e.g., mass balances). The Threshold Value for identified leakage is set at a higher value due to the lesser significance of identified leakage in comparison to unidentified or pressure boundary leakage.
- 1. 022G2.1.19 001/LOCT AND LOIT/SRO/M/F 3.9/3.8/022G2.1.19/LO-TA-63013///
Initial conditions:
- Unit 1 is at 100% reactor power.
- Containment Cooler
- 1 high speed fan breaker trips.
Currect conditions:
- ALB01-E06 CNM T HI TEMP is received. - IPC data for contai nment temperature is collected.
Which one of the following completes the following statement?
Based on the IPC data provided, the Te ch Spec Surveillance for containment temperature (Tech Spec SR 3.6.5.1) __(1)__ within Tech Spec limits, and per the applicable Annunciator Response Pr ocedure and System Operating Procedure, the crew is required, as a minimum, to __(2)__.REFERENCE PROVIDED (1) is (2) start one additiona l Containment Cooler (1) is (2) stop Containment Cooler #2, then start an additional pair of Containment Coolers (1) is NOT (2) start one additiona l Containment Cooler (1) is NOT (2) stop Containment Cooler #2, then start an additional pair of Containment Coolers A.B.C.D.K/A 022 Containment Cooling AA2.01 Ability to use plant computers to evaluate system or component status.K/A MATCH ANALYSISFriday, February 21, 2014 11:25:14 AM 1
The question sets up plausible scenario wher e the candidate must first determine which instruments are required to be referenced when complying with the Containment temperature monitoring verification surveill ance of Technical Specifications. A decision is required based on the operating limit in comparison with an IPC screen shot. The second part matches the KA for the Shift Su pervisor to evaluate Containment Cooling requirements used to control within envir onmental limits and the first part makes this SRO required knowledge when the surv eillance requirement is tested.EXPLANATION OF REQUIRED KNOWLEDGE Per TS SR 3.6.5.1, containment average temperature is required to be verified <120F on a periodic bases. The IP C screen for Control Room TS Rounds lists values for CMNT Levels 2, C, & B and an average te mperature. Individual readings above 120F are acceptable as long as th e average temper ature is <120F.
Per OSP 14000-1, page 15, if the IPC point s for containment temperature are not available, annunciator ALB01-E06 can be veri fied extinquished as an alternate. This annunciator utilzes temperatur e elements from all three le vels and has a setpoint of
<120F. If the annunciator is in alarm and the IPC is unavail able, then local temperature readings are required.
Per ARP 17001-1 for ALB01-E06, if any of the three contai nment temperature indicators rise above 120F, the operator is instructed to start an additional pair of Containment Coolers. Contai nment Coolers must be start ed in specific pairs (1&2, 3&4, 5&6, and 7&8) due to the backdrafter dampe r for the indivi dual fans being de-energized open on a commo n plentum (reference P&ID 1X4DB212 and SOP 13120-1 section 4.1). Starting the fans one at a time will re sult in the non-running fan of the pair to spinning backwards. When the second fan is subsequently started, the supply breaker will trip ope n from high in-rush current produced by the increasedelectrical slip angle.
Since fan #1 tripped, fan #2 s hould be stopped prior to star ting an additional pair. Fan#2 is mostly recirculating short-cycled flow in the current configuration and is producing little cooling to containment. Simply st arting a second fan on another pair would only create the same condition on a second pair and would also do little to lower containment temperature.ANSWER / DISTRACTOR ANALYSISA. Incorrect. Plausible. Part 1 is correct and sets up a circumstance were the candidate must determine which instrumentat ion in particular is required to be monitored the ensure compliance with SR 3.6.5.1 for
Containment temperature. T he SR requires the average temperature of three levels in Containment be noted to ensure
Containment temperature remain s within the accident analyses.Part 2 is incorrect but 'plausible' because the candidate may determine that the onl y required action is to start an additional Containment Cooler to replac e the tripped Containment coolernot recalling that the system operating procedure would requireFriday, February 21, 2014 11:25:14 AM 2
the fans be operated in pairs.B. Correct. Part 1 is correct.
See Part 1 of choice A above.
Part 2 is correct per the ARP 17001-1, and the SOP 13120-1, an additional pair of Containm ent Coolers would be started.
C. Incorrect. Plausible. Part 1 is incorrect but 'plausible' in that the candidate may determine that Technical Specificat ion if any of the three levels listed on the IPC printout exceeded 120F. This assumption would be consistent with the us e of ALB01-E06 as an alternate as described in OSP 14000-1.
Part 2 is incorrect. See Part 2 of choice A above.
D. Incorrect. Plausible. Part 1 is in correct. See Part 1 of choice C above.
Part 2 is correct. See part 2 of choice B above.SRO JUSTIFICATION(2) Facility operating limitations in the technical specifications and their bases.-Can question be answered solely by knowing = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action?
No, the question does not address 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Tech Spec actions.-Can question be answered solely by knowing the LCO/TRM information listed "above-the-line?"
No, the question is not related to above-the-line information in Tech Spec.
-Can question be answered solely by knowing the TS Safety Limits?
No, thequestion is not related the Tech Spec Safety Limits.-Does the question involve one or more of the following for TS,TRM, or ODCM?* Application of Required Acti ons (Section 3) and SurveillanceRequirements (Section 4) in accordance with rules of applicationrequirements (Section 1).
Yes, the question requires specific knowledgerelated to Tech Spec surveillance re quirements and instruments used tosatisfy the requirement.* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4).
No, the question is no t related to these aspects of Tech Spec.* Knowledge of TS bases that is required to analyze TS required actions and terminology.
No, the question is not related to Tech Spec bases.Friday, February 21, 2014 11:25:14 AM 3
Level: SROTier # / Group # T2 / G1 K/A# 022G2.1.19 Importance Rating: 3.9 / 3.8 Technical
Reference:
TS 3.6.5, Amendment No. 158, page 3.6.5-1 TS 3.6.5 Bases Rev 14.0, page 3.6.5-4
SOP 13120-1, Rev 24.0, pages 3 & 5-8
ARP 17001-1, Rev 31.1, page 3 & 47
OSP 14000-1, Rev 88.2, page 15
P&ID 1X4DB212, Rev 12.0References provided: IPC Screen shot of CNMT Temps Learning Objective:
LO-TA-63013 Implement Tec hnical Specification LCO using 10008-C (SRO Only)
LO-LP-39209-01 For any given it em in section 3.5 of Tech Specs, be able to: St ate the LCO. State any one hour or less required actions LO-LP-39209-03Describe the bases for any given Tech Spec in section 3.5.LO-PP-29101-15 State the starting interlocks associatedwith the Containment Cooling fans.
Include set points and coincidence where applicable.Question origin: MODI FIED - LORQ Question
- V-LO-PP-29101-09 001Cognitive Level: M/F 10 CFR Part 55 Content: 41.9 / 43.2 Comments: You have completed the test!Friday, February 21, 2014 11:25:14 AM 4
- 1. V-LO-PP-29101-09 001/SRO/
0022A2.04/LO-TA-29003/C/A/3/DIABLO CANYON 05/SOP 13120-1/2/TECH SPEC 3.6.5 Due to the high summer air temperature and fouling of the Contai nment Air Coolers are causing the following conditions to occur on Unit 1: - ALB01-E06 "CNMT HI TEMP" is in alarm - Containment Level 2 temperature - 122°F - Containment Level C temperature - 114°F - Containment Level B temperature - 120°F - Containment Coolers 1,2,5, and 6 are running in high speed - Containment pressu re is currently 1.0 ps ig and increasing slowly Based on the current plant conditions, which of the follo wing actions should the Shift Supervisor the operator to perform?
Direct a start of the Train "B" Containmen t Coolers in Hi speed, and consider venting Containment.
Direct a start of Train "B" Containment Coolers in Hi speed, and restore pressure within limits within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in MODE 3 in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Direct a start of the Train "B" Containmen t Coolers in Hi speed, reduce Containment temperature to less than 120°F within the nex t 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in MODE 3 in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and consider venting Containment.
Direct a start of the Train "B" Containment Coolers in Hi speed, restore pressure within limits within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in MODE 3 in the next 6 hrs and consider venting Containment.
A.B.C.D.Monday, January 20, 2014 11:59:42 AM 1
Containment Air Temperature
3.6.5 Vogtle
Units 1 and 2 3.6.5-1 Amendment No. 158 (Unit 1) Amendment No. 140 (Unit 2) 3.6 CONTAINMENT SYSTEMS
3.6.5 Containment
Air Temperature
LCO 3.6.5 Containment average air temperature shall be 120°F.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Containment average air temperature not within limit. A.1 Restore containment average air temperature to within limit.
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> B. Required Action and associated Completion Time not met.
B.1 Be in MODE 3.
AND B.2 Be in MODE 5.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.1 Verify containment average air temperature is within limit.
In accordance with the Surveillance Frequency Control Program
Containment Air Temperature B 3.6.5 BASES (continued)
SURVEILLANCE SR 3.6.5.1 REQUIREMENTS Location Tag Number
- a. Level 2 TE-2563 b. Level B TE-2613
- c. Level C TE-2612 NOTE: A local sample may be taken at a corresponding location in lieu of using one of the instruments designated above.
Verifying that containment average air temperature is within the LCO
limit ensures that containment operation remains within the limit assumed for the containment analyses. In order to determine the containment average air temperature, an arithmetic average is calculated using measurements taken at locations within the containment selected to provide a representative sample of the overall containment atmosphere. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES 1. FSAR, Section 6.2.
- 2. 10 CFR 50.49.
Approved By M. D. Askew Vogtle Electric Generating Plant Procedure Number Rev 13120-1 24 Date Approved 01/15/2012 CONTAINMENT BUILDING COOLING SYSTEM Page Number 3 of 50 Printed February 21, 2014 at 10:34
1.0 PURPOSE
This procedure provides instructions for operation of the Containment Building Cooling Systems, which consist of t hese subsystems: Containment Heat Removal System (CHRS), Control Rod Dr ive Mechanism (CRDM) Cooling Units, and the Containment Building (CTB) Cavity Cooling System. Instructions are provided in the following subsections:
4.1 CTB Cooling System Startup To Standby
4.2 Containment
Heat Re moval System Startup
4.3 CRDM Cooling Units Startup
4.4 Shifting
CRDM Cooling Units
4.5 CTB Cavity Cooling System Startup
4.6 Shifting
CTB Cavity Cooling Units
4.7 Shifting
CTB Reactor Support Cooling Fans
4.8 Shifting
CTB Cooling Unit Fans
4.9 Shifting
Auxiliary Coolers
4.10 Post LOCA Purge Cavity Fan Operation
4.11 Containment Heat Re moval System Shutdown
4.12 CRDM Cooling Units Shutdown
4.13 CTB Cavity Cooling System Shutdown
Approved By M. D. Askew Vogtle Electric Generating Plant Procedure Number Rev 13120-1 24 Date Approved 01/15/2012 CONTAINMENT BUILDING COOLING SYSTEM Page Number 5 of 50 INITIALS Printed February 21, 2014 at 10:34
4.0 INSTRUCTIONS
4.1 CTB COOLING SYSTEM STARTUP TO STANDBY
4.1.1 Perform
Table 1 to align the system handswitches for startup. ________
4.1.2 If required, perform the system startup a lignment per 11120-1, "Containment Building Cooling System Alignment". ________
4.1.3 Close
the links for the K2 Relay and close the breakers for the CNMT COOLING UNITS (IV REQUIRED), document on Checklist 1:
1ABE-26 ________
1ABE-27 ________
1ABC-07 ________
1ABC-08 ________
1BBE-26 ________
1BBE-27 ________
1BBC-07 ________
1BBC-08 ________
Approved By M. D. Askew Vogtle Electric Generating Plant Procedure Number Rev 13120-1 24 Date Approved 01/15/2012 CONTAINMENT BUILDING COOLING SYSTEM Page Number 6 of 50 INITIALS Printed February 21, 2014 at 10:34 NOTE Panel LR01 (1-1816-U3-019) is located in Control Building, Corridor R-149, outside the Radiochemistry Lab.
4.1.4 Open the CTB Cooling Units Outlet Dampers at Panel LR01.
Damper Position may also be verified by 1-ZLB indications on
1-QHVC. (IV REQUIRED), IF ava ilable, the IPC may also be used to verify damper position, document on Checklist 1.
DAMPER DESCRIPTION HANDSWITCH 1-QHVC 1-HV-2582A CTB CLG UNIT 1 1-HS-2582G 1ZLB-43 ________
1-HV-2582B CTB CLG UNIT 2 1-HS-2582H 1ZLB-43 ________
1-HV-2583A CTB CLG UNIT 3 1-HS-2583G 1ZLB-44 ________
1-HV-2583B CTB CLG UNIT 4 1-HS-2583H 1ZLB-44 ________
1-HV-2584A CTB CLG UNIT 5 1-HS-2584G 1ZLB-43 ________
1-HV-2584B CTB CLG UNIT 6 1-HS-2584H 1ZLB-43 ________
1-HV-2585A CTB CLG UNIT 7 1-HS-2585G 1ZLB-44 ________
1-HV-2585B CTB CLG UNIT 8 1-HS-2585H 1ZLB-44 ________
Approved By M. D. Askew Vogtle Electric Generating Plant Procedure Number Rev 13120-1 24 Date Approved 01/15/2012 CONTAINMENT BUILDING COOLING SYSTEM Page Number 7 of 50 INITIALS Printed February 21, 2014 at 10:34 4.1.5 Open the links for the K2 Relay for the following breakers. (IV REQUIRED), document on Checklist 1:
1ABE-26 ________
1ABE-27 ________
1ABC-07 ________
1ABC-08 ________
1BBE-26 ________
1BBE-27 ________
1BBC-07 ________
1BBC-08 ________
Approved By M. D. Askew Vogtle Electric Generating Plant Procedure Number Rev 13120-1 24 Date Approved 01/15/2012 CONTAINMENT BUILDING COOLING SYSTEM Page Number 8 of 50 INITIALS Printed February 21, 2014 at 10:34 NOTES CTB Cooling Units Outlet Dampers are OPEN and DE-ENERGIZED with their breakers LOCKED OPEN to prec lude inadvertent closure of dampers. Damper indication on 1ZLB-43 and 1ZLB-44 will have no indication after performing the following step. Indication will still be available for dam per indication on the IPC. Computer points for damper indication are; ZD9430, ZD9432, ZD9434, ZD9436, ZD9438, ZD9440, ZD9442 and ZD9444.
4.1.6 Open and lock the following breakers for the CTB Cooling Units Outlet Dampers. (IV REQUIRED), document on Checklist 1:
DESCRIPTION BREAKER CTB CLG UNIT A7-001 1-HV-2582A, 1ABE-26 ________
CTB CLG UNIT A7-002 1-HV-2582B, 1ABE-27 ________
CTB CLG UNIT A7-003 1-HV-2583A, 1BBE-26 ________
CTB CLG UNIT A7-004 1-HV-2583B, 1BBE-27 ________
CTB CLG UNIT A7-005 1-HV-2584A, 1ABC-07 ________
CTB CLG UNIT A7-006 1-HV-2584B, 1ABC-08 ________
CTB CLG UNIT A7-007 1-HV-2585A, 1BBC-07 ________
CTB CLG UNIT A7-008 1-HV-2585B, 1BBC-08 ________
Approved By S. E. Prewitt Vogtle Electric Generating Plant Procedure Number Rev 17001-1 31.1 Date Approved 08/16/2010 ANNUNCIATOR RESPONSE PROCE DURES FOR ALB 01 ON PANEL 1A1 ON MCB Page Number 3 of 48 Printed October 2, 2013 at 9:13 (1) (2) (3) (4) (5) (6)
A CIRC WTR P-1 MOTOR OVERLOAD
CIRC WTR P-2 MOTOR OVERLOAD TPCW PUMP 1 TRIPPED TPCW PUMP 2 TRIPPED SERVICE AIR SWING CMPSR
MISALIGNED
AIR CMPSR
B CIRC WTR P-1 DISCH VLV TROUBLE
CIRC WTR P-2 DISCH VLV TROUBLE TPCW PMP DISCH HDR LO PRESS SERVICE AIR
CMPSR TROUBLE INSTR AIR EQUIP LO PRESS
C CIRC WTR P-1
LO PIT LVL
CIRC WTR P-2
LO PIT LVL
CLG TOWER
BASIN HI LVL
UNIT 1 SERV AIR HDR
TIED TO UNIT 2
SERVICE AIR
D CIRC WTR P-1
SCREEN WTR
HI DIFF LVL
CIRC WTR P-2
SCREEN WTR
HI DIFF LVL
TPCCW PUMP 1 TRIPPED TPCCW PUMP 2 TRIPPED INSTR AIR CNMT SPLY
LINE BREAK
E CONDR CIRC WTR ISO VLV CLOSED
TPCCW SURGE TK HI/LO LVL
TPCCW DISCH HDR LO PRESS CNMT HI TEMP
F
CNMT HI MSTR Approved By S. E. Prewitt Vogtle Electric Generating Plant Procedure Number Rev 17001-1 31.1 Date Approved 08/16/2010 ANNUNCIATOR RESPONSE PROCE DURES FOR ALB 01 ON PANEL 1A1 ON MCB Page Number 47 of 48 Printed October 2, 2013 at 9:13 WINDOW E06
ORIGIN SETPOINT 1-TSH-2563 1-TSH-2612
1-TSH-2613
120 F
CNMT HI TEMP
1.0 PROBABLE
CAUSE Insufficient number of Containment Building Cooling Units operating.
2.0 AUTOMATIC
ACTIONS NONE
3.0 INITIAL
OPERATOR ACTIONS NONE
4.0 SUBSEQUENT
OPERATOR ACTIONS
- 1. Start an additional pair of Containment Cooling Units or a Containment Auxiliary Cooling Unit per 13120-1, "Containment Building Cooling Systems".
- 2. Verify Nuclear Service Cooling Water flow to coolers, and IF necessary, dispatch an operator to inspect the Cont ainment Heat Removal System.
- 3. Refer to Technical Specificati on LCO 3.6.5 and 3.6.6.
- 4. IF equipment failure is indicated, initiate maintenance as required.
5.0 COMPENSATORYOPERATOR
ACTIONS NONE
END OF SUB-PROCEDURE
REFERENCES:
1X4DB212, CX5DT101-66, CX5DT101-71 Approved By J.B. Stanley Vogtle Electric Generating Plant Procedure Version 14000-1 88.2 Effective Date 09/25/2013 OPERATIONS SHIFT AND DAILY SURVEILLANCE LOGS Page Number 15 of 36 Printed February 21, 2014 at 11:05 Sheet 9 of 10 DATA SHEET 1 MODE 1 & 2 MODE _______________
DATE _______________
LCO TECH SPEC INDICATION LIMIT(S)
METHOD OF VERIFICATION SURV REQ PARAMETER INSTRUMENT DAY NIGHT TOLERANCE LCO/PROC CREFS ACTUATION OPERABLE SR 3.3.7.1
FCN 3 CR INTAKE RADIATION 1RE-12116 CHANNEL CHECK
3.3.7 CHANNEL
CHECK MONITORS (INIT) 1RE-12117 REQUIRED 2 FHB ACTUATION OPERABLE TRS 13.3.6.1 FHB EFFL RADIOGAS ARE-2532A
- 13.3.6 CHANNEL CHECK FHB ISO (INIT) ARE-2532B REQUIRED 1 FHB ACTUATION OPERABLE TRS 13.3.6.1 FHB EFFL RADIOGAS ARE-2533A
- 13.3.6 CHANNEL CHECK FHB ISO (INIT) ARE-2533B REQUIRED 1 *INDICATING NORMALLY. ALL STATUS AND ALARM LIGHTS EXTINGUISHED. DG1A FUEL OIL INVENTORY VERIFY FUEL OIL STORAGE TANK LEVEL SR 3.8.3.1 DG 1A LEVEL (%) 1-LI-9024 82% 3.8.3 DG1B FUEL OIL INVENTORY VERIFY FUEL OIL STORAGE TANK LEVEL SR 3.8.3.1 DG 1B LEVEL (%) 1-LI-9025 82% 3.8.3 TWO INDEPENDENT CONTROL ROOM EMERGENCY FILTRATION SYSTEMS SHALL BE OPERABLE VERIFY CONTROL ROOM TEMP SR 3.7.10.1 SR 3.7.11.1
NOTE: TEMPERATURE INDICATION IS OBTAINED FROM HAND-HELD TEST EQUIPMENT. RECORD INSTRUMENT INFORMATION BELOW.
INSTRUMENT ID NO.
N/A CAL DUE DATE CONTROL ROOM TEMPERATURE (F) M&TE <85 F 3.7.10 3.7.11 THE RWST SHALL BE OPERABLE VERIFY TEMPERATURE SR 3.5.4.1 TRS 13.1.7.1 RWST TEMPERATURE (F) 1TIS-10980
>51 F * <109 F
- 3.5.4 13.1.7 *WITH INDICATED RWST TEMPERATURE OUTSIDE THE LIMITS, THEN VERIFY RWST TEMPERATURE IS WITHIN TECHNICAL SPECIFICATION LIMITS BY PLACING THE RWST ON RECIRC USING SLUDGE MIXING PUMP WITH HEATER OFF AND OBSERVING 1-TI-10982 TO BE WITHIN 44F AND 116F. THE ULTIMATE HEAT SINK SHALL BE OPERABLE COMPUTER POINT T2601* <90 F 3.7.9 VERIFY WATER -OR- TEMPERATURE AND LEVEL SR 3.7.9.2 TEMPERATURE (F) 1TJI-1692 POINT 2* COMPUTER POINT T2602* -OR- 1TJI-1692 POINT 17* *IF COMPUTER POINT AND RECORDER POINT ARE NOT AVAILABLE, TEMPERATURE READING MUST BE OBTAINED LOCALLY USING HAND-HELD TEST EQUIPMENT. RECORD INSTRUMENT INFORMATION BELOW.
INSTRUMENT ID NO.
N/A CAL DUE DATE SR 3.7.9.1 LEVEL 1LI-1606 >73% (%) 1LI-1607 CONTAINMENT AIR TEMPERATURE SHALL NOT SR 3.6.5.1 COMPUTER POINT T2501 EXCEED 120F VERIFY AVERAGE AIR TEMPERATURE (F) COMPUTER POINT T2502 NA TEMPERATURE COMPUTER POINT T2503 COMPUTER POINT UT2501 (AVG)
<120 F 3.6.5 *IF COMPUTER POINT IS NOT AVAILABLE VERIFY CNMT HI TEMP ALARM ALB-01 (E06) IS NOT IN ALARM. ALB-01 (E06) NOT IN ALARM
- IF COMPUTER POINT AND ALB-01 (E06) ARE NOT AVAILABLE, TEMPERATURE READING MUST BE OBTAINED LOCALLY USING HAND-HELD TEST EQUIPMENT FOR 1TE-2612 FOR POINT T2502 AND 1TE-2613. FOR POINT T2503 RECORD INSTRUMENT INFORMATION BELOW. USE MCB INDICATOR 1TI-2563 FOR POINT T2501 AND AVERAGE THE THREE.
INSTRUMENT ID NO.
<120 F CAL DUE DATE COMPLETED BY: DAY: TIME: NIGHT: TIME: SS REVIEW: DAY: TIME: NIGHT: TIME:
- 1. 025AG2.4.30 001/LOCT AND LOIT/SRO/C/A 2.7/4.1/025AG2.4.30/LO-TA-40002///At time 1055:
- Unit 1 is in Mode 5.
- Containment in tegrity is NOT established. - Pressurizer level is 27%.
- All Pressurizer Safety Valves are removed.At time 1100: - An LOSP occurs.
- DG1A trips on overspeed.
- RHR pump 'B' will NOT st art following t he load sequence. - RCS temperatur e is 200°F and increasing.
Which one of the following completes the following statement?At time 1115 , the Shift Manager must declare a(n) __(1)__ per NMP-EP-110,"Emergency Classification Determ ination and Initial Action," and no later than time 1130 , as a minimum, the __(2)__ must be notified of the declaration per NMP-EP-111, "Emergency Notifications."REFERENCE PROVIDED
__(1)__ __(2)__ NOUE NRC, state, and local authorities NOUE state and local authorities Alert NRC, state, and local authorities Alert state and local authorities A.B.C.D.K/A 025 Loss of RHR SystemG2.4.30 Knowledge of events related to syst em operation/status that must bereported to internal organizations or external agenci es, such as the State, the NRC, or the transmission system operator.K/A MATCH ANALYSIS The question sets up a plausib le scenario which includes th e first required element ofFriday, February 21, 2014 12:40:33 PM 1
the KA loss of RHR co oling, then has the candidate ev aluate plant conditions and make determination of repor ting requirements which meets t he second part of the KA and brings the knowledge to the SRO required level.EXPLANATION OF REQUIRED KNOWLEDGE Based on the conditions in the stem, 2 EPI Ps classification th resholds have been exceed. First, an LOSP has existed with only the 'B' DG energizi ng 1BA03. As such NOUE CU3 has been exceeded.
Additonally, due to both RHR pumps not running, RCS temperature has risen above 200F while in Mode 4. Containmen t integrity is not established. Since all pressurizer safeti es are removed, RCS integrity is also not established. As shuch, ALERT CA4 has also been ex ceeded. Therefore, the Emergency Director is requir ed to declare an ALERT emerg ency on or before 11:15 per NMP-EP-110.
Per NMP-EP-111 steps 5.1.1 ad 5.1.3, state and local agenc ies must be notified within 15 minutes of the declaration, which woul d be 11:30. Additionally, the NRC shall be notified immediately follwoi ng the state and local agencies and within an hour of the declaration.
ANSWER / DISTRACTOR ANALYSISA. Incorrect. Plausible. Part 1 is inco rrect but plausible bec ause the candidate may determine the loss of power ev ent is driving the initial classification not recognizing the loss of RHR Cooling and CU3 would be the correct.
Part 2 is incorrect but 'plausible' because the candidate may determine that the NRC must be not ified in this condition within 15 minutes. Per NMP-EP-111, Noti fication of the NRC shall be completed immediately following notifications to the state and local agencies and with in an hour of the declaration of an emergency. Therefore, the NRC is not REQUIRED to be notified by 11:30, but instead is allowed additonal time if needed up to an hour, but is to be do ne as soon as possible following the state and locals.B. Incorrect. Plausible. Part 1 is incorrect. See Part 1 of choice A above.
Part 2 is correct in that within 15 minutes of the classification of the NOUE the state and local authorities must be notified per NMP-EP-111, "Emergency Notifi cations". Notifications of applicable State and Local Agenc ies shall be accomplished as soon as practicable and within 15 minutes of the declaration of an emergency, an upgrade to a hi gher emergency classification level, or the approval of prot ective actions recommendations.C. Incorrect. Plausible. Part 1 is co rrect and has the cand idate evaluate plant conditions and determine the event meets the ALERT threshold for CA4 due to being in Mode 5 with RCS temperature >200F without containment and RCS integrity established.Friday, February 21, 2014 12:40:33 PM 2
without containment and RCS integrity established.
Part 2 is incorrect. See Part 2 of choice A above.D. Correct. Part 1 is correct.
See Part 1 of choice C above.
Part 2 is correct. See Part 2 of choice B above.SRO JUSTIFICATION(5) Assessment of facility conditions and selection of appropriate proceduresduring normal, abnormal, and emergency situations.-Can the question be answered solely by knowing "systems knowledge", i.e.,how the system works, flowpath, logic, component location?
No, the answerrequires specific knowledge of emergency classification thresholds.-Can the question be answered solely by knowing immediate operator actions? No, IOAs are not addressed by this question.
-Can the question be answered solely by knowing entry conditions for AOPs orplant parameters that require direct entry to major EOPs?
No, the question does not address AOP or EOP entry conditions.-Can the question be answered solely by knowing the purpose, overall sequenceof events, or overall mitigative strategy of a procedure?
No, the answer requiresspecific knowledge of emergency classification thresholds.-Does the question re quire one or more of the following?* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or se ction of a procedure to mi tigate, recover, or with which to proceed* Knowledge of when to impl ement attachments and appendices, including how to coordinate these it ems with procedure steps* Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific s ubprocedures or emergency contingency procedures* Knowledge of administrative procedures that specify hierarchy,implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
Yes, the answer requires specific knowledge of emergency classification thresholds and determin ation of the specific classification based on current plant conditions. This is an SRO ONLYjob link associated with an SRO ONLY objecti ve. [LO-LP-40101-13 Givenan emergency scenario, and the pro cedure, classify the emergency (SROonly).]Friday, February 21, 2014 12:40:33 PM 3
Level: SROTier # / Group # T1 / G1 K/A# 025G2.4.30 Importance Rating: 2.7 / 4.1 Technical
Reference:
NMP-EP-110-GL03, Rev 3.0, page 58 NMP-EP-110-GL03, Figure 3, Rev 3.0, page 123 NMP-EP-111, Re v 8.0, page 7 References provided:
NMP-EP-110-GL03, Figure 1, Rev 3.0, page 121 NMP-EP-110-GL03, Figure 2, Rev 3.0, page 122 NMP-EP-110-GL03, Figure 3, Rev 3.0, page 123Learning Objective: LO-TA-40002 Emergency Classification andImplementing Instructions using
NMP-EP-110 (SRO Only)LO-TA-40003 Emergency Notifications using NMP-EP-111 LO-LP-40101-13 Given an em ergency scenario, and the procedure, classify the emergency (SRO
only).LO-LP-40101-16 List the stat e and federal authorities that are notified in an emergency.Question origin: NEW Cognitive Level: C/A 10 CFR Part 55 Content: 43.5 Comments: You have completed the test!Friday, February 21, 2014 12:40:33 PM 4
NMP-EP-110-GL03 - VEGP EALs - ICs, Threshold Values and Basis Version 3.0 123 FIGURE 3 Southern Nuclear Operating Company Emergency Implementing Procedure Emergency Notifications NMP-EP-111 Version 8.0 Page 7 of 12 4.3.3 The ENS functions DO NOT normally transfer to the EOF. The EOF has an ENS communicator position that coordinates with the TSC ENS communicator to maintain a continuous open communication line with the NRC. The role of the EOF ENS Communicator is to assist in communications with the NRCOC relevant to activities performed from the EOF (i.e., offsite interface, public information, PAR development, Dose assessment activities, etc.).
5.0 PROCEDURE
5.1 Precautions and Limitations 5.1.1 Notifications of applicable State and Local Agencies shall be accomplished as soon as practicable and within 15 minutes of the declaration of an emergency, an upgrade to a higher emergency classification level, or the approval of protective actions recommendations. 5.1.2 Electronic notification using the electronic Emergency Notification Form (ENF) in WebEOC is the preferred method of notification of State and Local Agencies. Should WebEOC be unavailable the back-up notification method is completion of a hard copy ENF and reading the form to the applicable State and Local Agencies via the ENN. Whether using the electronic or back-up ENF the emergency notification to an agency is considered complete when the agency verbally confirms receipt of the message via the ENN. To expedite availability of WebEOC in an emergency, the crew members responsible for completing the ENF and making electronic notifications should login to WebEOC as soon as possible and remain logged-in. 5.1.3 Notification of the NRC shall be completed immediately following notifications to the state and county agencies and within an hour of the declaration of an emergency. Follow-up notifications of the NRC shall be made promptly after any further degradation in the plant conditions, any change from one emergency class to another, or for the termination of an emergency. NRC notifications are typi cally performed utilizing the Federal Telephone System (FTS). The Emergency Notification System (ENS) line is normally utilized. An open line is maintained for the duration of the event at the request of the NRC communicator receiving the initial notification. 5.1.4 For security based emergencies, notifications to the NRC should be performed within 15 minutes of discovery of an imminent threat or attack against the plant to ensure proper mobilization of federal resources.
5.1.5 If the plant condition degrades and a higher emergency classification is declared before the notifications are confirmed for the lesser emergency declaration, then a notification reflecting the higher emergency classification should be made. This notification should be made within 15 minutes of the lesser emergency declaration. This should be performed IF the notification can be made within 15 minutes of the lesser (first) classification. CAUTION An initial notification of an upgrade in emergency Classification should take precedence over a follow-up message of a lower ranking emergency. (i.e., an initial site area emergency notification takes precedence over an alert follow-up notification.
)
- 1. 026AA2.06 001/LOIT AND LOCT/SRO/C/A 2.8/3.1/026AA2.06/LO-TA-60005///At time 1000:
- Unit 1 is at 100% reactor power. At time 1005: - ALB04-A03 ACCW RCP 1 CLR LOW FLOW is received. - ALB04-A04 ACCW RCP 1 CLR OUTLET HI TEMP is received.
- 18022-C, "Loss of Auxiliary Component Cooling Water," is entered.At time 1007: - RCP #1 seal water inlet temper ature is 220°F and ri sing at 1°F per minute.
- RCP #1 motor stator winding temper ature is 307°F and rising at 1°F per minute.
Which one of the following completes the following statement?
To prevent damage to the RCP , the RCP must be stopped no later than time __(1)__, and after the reactor is tripped, t he Shift Supervisor directs th e OATC to stop the affectedRCP per __(2)__ direction.
(1) 1012 (2) 18022-C, "Loss of Auxiliary Component Cooling Water"
(1) 1012 (2) 19000-C, "Reactor Trip or Safety Injection" (1) 1016 (2) 18022-C, "Loss of Aux iliary Component Cooling Water" (1) 1016 (2) 19000-C, "Reactor Trip or Safety Injection" A.B.C.D.K/A 026 Loss of Component Cooling Water AA2.06 Ability to determine and interpret the following as they apply to theLoss of Component Cooling Water: - The length of time after the loss of CCW flow to a componentFriday, February 21, 2014 1:21:51 PM 1
before that component may be damaged.K/A MATCH ANALYSIS To answer this question, the applicant must kn ow that loss of ACCW will lead to entry into 19000-C, requiring a m anual trip (even though it d oes not approach or exceed any automatic setpoints), and whic h procedure provides the action. Since the Shift Supervisor is required to recall the specif ic strategy from t he 18022-C procedure, the choices involve SRO only kno wledge of the procedure flow path of the AOP (past the entry conditions and there are no immediate actions in 180 22-C), and knowledge of the need to enter 19000-C. In addition, the question requires the c andidate to recall not only how long the RCPs can operate without ACCW coolin g but to evaluate other parameters that may require i mmediate stopping of the pumps.EXPLANATION OF REQUIRED KNOWLEDGE At 1005, annunciators ALB04-A03 and A04 indicate a co mplete loss of ACCW flow toRCP #1 only. AOP 18022-C is entered. Step 6 checks to see if the RCP should be stopped. At 1016, RCP #1 wo uld have sustained a total loss of ACCW for >10 minutes and must be stopped. The stem al so states that seal water inlet temperature is rising at a rate of 1F/min. At 1018, the RCP must be stopped due to seal water inlet temperature. Additionally, motor stator winding temperature is al so rising at a rate f 1F/min. At 1012, 311F would have been exceeded and the RCP must be stopped.
Therefore, the most limi ting operationg c ondition would be the stator winding temperature and the the RCP must be stopped no later than 1012. The RCP operationg parameters in 18022-C are the same as those listed in SOP 13003-1 Limitation 2.2.10.
Since the reactor is greater than 15% RTP, the reactor must be tripped prior to tripping the RCP. Tripping the reactor is a direct entry condition into EOP 19000-C. On step 11 of 19000-C, a check of ACCW status is made and the RNO directs stopping the RCP.
However, 18022 step 6 is specifically design to trip the reactor and stop the RCP prior to performing EOP 19000-C. Th is is done to ensure the RCP is not damaged in the time required to perform the in itial steps of 19000-C before direction to stop the RCP is encountered.ANSWER / DISTRACTOR ANALYSISA. Correct. Part 1 is correct. Base d of Westinghouse ve ndor requirements the RCPs must be stopped under normal conditions when ACCW cooling is lost to the motors to prevent potential damage. This is supported by SOP 13003-1, which establishes the ACCW operating ti me limit of 10 minutes and other trip parameters. In the case pres ented to the candidate the most limiting parameter is the stator limit of 311F which will require the pump to be stopped within 5 minutes of 10:07 (i.e. 10:12).
Part 2 is correct because 18022-C directs manual reactor trip, verify trip, the stop RCPs th en perform 19000-C, therefore the correct direction to stop the RCP is 18022-C as opposed to 19000-C.Friday, February 21, 2014 1:21:51 PM 2
B. Incorrect. Plausible. Part 1 is correct. See Part 1 of choice A above. Part 2 is incorrect but 'plausible' because the candidate may
determine that the direction that is provided in 19000-C step 11 would be used to stop the RCPs as opposed to 18022-C step 1.
The 10 minute time allowance could lead the candidate to
determine there is no real pressu re to speed up the action to stop the RCPs, prior to the initial steps in 19000-C. C. Incorrect. Plausible. Part 1 is incorr ect however is 'plausib le' since the candidate may determine that none of the operating limits provided in the stem is above limits or will reac h their limit within 10 minutes, therefore the 10 minutes becomes the most limiting factor.
Part 2 is correct. See Part 2 of choice A above.
D. Incorrect. Plausible. Part 1 is in correct. See Part 1 of choice C above.
Part 2 is correct. See Part 2 of choice B above.SRO JUSTIFICATION(5) Assessment of facility conditions and selection of appropriate proceduresduring normal, abnormal, and emergency situations.-Can the question be answered solely by knowing "systems knowledge", i.e.,how the system works, flowpath, logic, component location?
No, the systemparameter needed to determine if the RCP should be stopped is system levelknowledge. However, the direction to stop the RCP requires specific knowledge and priortization of steps in both the AOP and EOP.-Can the question be answered solely by knowing immediate operator actions? No, there are no IOAs associated with the actions addressed in this question.-Can the question be answered solely by knowing entry conditions for AOPs orplant parameters that require direct entry to major EOPs?
No, stopping an RCPhas nothing to do with the entry conditions for either AOP 18022-C or EOP 19000-C. 19000-C does have entry conditions associated with trip of the reactor,which would stem from stopping an RCP at 100% RTP. However, this would leada candidate down the incorrect path.
-Can the question be answered solely by knowing the purpose, overall sequenceof events, or overall mitigative strategy of a procedure?
No, specific knowledgeof individual procedur e steps is required.-Does the question re quire one or more of the following?* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or se ction of a procedure to mi tigate, recover, or with which to proceed
- Knowledge of when to impl ement attachments and appendices, including how to coordinate these it ems with procedure steps* Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific subprocedures or emergencyFriday, February 21, 2014 1:21:51 PM 3
contingency procedures Yes, knowledge of specific diagnotics in step 6 of AOP 18022-C is requir ed in contrast to step 11 of EOP 19000-C.* Knowledge of administrative proc edures that specify hierarchy, implementation, and/or co ordination of plant norma l, abnormal, and emergency procedures Level: SROTier # / Group # T1 / G1 K/A# 026AA2.06 Importance Rating: 2.8 / 3.1Technical
Reference:
EO P 19000-C Rev 37.1, page 21 AOP 18022-C Rev 15.2, page 6
SOP 13003-1, Rev 47.1, page 7References provided: None Learning Objective: LO-TA-60005 Respond to a Loss of ACCW per 18022-C LO-LP-60318-05 Descr ibe the operator actions required during a loss of ACCW with the plant in
operation and the RCP temperature time limits are exceeded.
LO-PP-16401-07 List the RCP components that are cooledby the ACCW system.Question origin: NEW Cognitive Level: C/A 10 CFR Part 55 Content: 43.5 Comments: You have completed the test!Friday, February 21, 2014 1:21:51 PM 4
Approved By Vogtle Electric Generating Plant Procedure Version M.G. Brill 19000-C 37.1 Effective Date E-0 REACTOR TRIP OR SAFETY INJECTION Page Number 7-5-13 21 of 35 OATC INITIAL ACTIONS Sheet 5 of 4 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED Printed October 24, 2013 at 09:00 4 11. Check ACCW Pumps - AT LEAST ONE RUNNING.
- 11. Try to start one ACCW Pump.
11 IF an ACCW Pump can NOT be started within 10 minutes of loss of ACCW, THEN stop all RCPs.
IF an ACCW Pump can NOT be started within 30 minutes of loss of ACCW, THEN close ACCW Containment isolation valves:
ACCW SPLY HDR ORC ISO VLV HV-1979 ACCW SPLY HDR IRC ISO VLV HV-1978 ACCW RTN HDR IRC ISO VLV HV-1974 ACCW RTN HDR ORC ISO VLV HV-1975
- 12. Adjust Seal Injection flow to all RCPs 8 TO 13 GPM.
- 12. 12 13. Dispatch Operator to ensure one train of SPENT FUEL POOL COOLING in
service per 13719, SPENT FUEL
POOL COOLING AND
PURIFICATION SYSTEM.
- 13. IF one train of SFP COOLING can NOT be restored to service, THEN initiate 18030-C, LOSS OF SPENT FUEL POOL LEVEL OR COOLING. 13 END OF SUB-PROCEDURE TEXT
Approved By Vogtle Electric Generating Plant Procedure Version J. B. Stanley 18022-C 15.2 Effective Date LOSS OF AUXILIARY COMPONENT COOLING WATER Page Number 05/07/2013 6 of 13 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED Printed January 20, 2014 at 12:46
- 6. Check if RCPs should be stopped:
- 6. 6 a. Check the following RCP parameters (using plant
computer):
- a. Perform the following:
6.a 1) IF any parameter limit is exceeded, THEN perform Step 6.b.
6.a.1) Motor bearing (upper or lower radial or thrust) -
GREATER THAN 195°F.
- 2) Go to Step 7.
6.a.2) Motor stator winding - GREATER THAN 311°F.
Seal water inlet - GREATER THAN 230°F.
Loss of ACCW - GREATER THAN 10 MINUTES.
- b. Perform the following:
- b. 6.b 1) Trip the reactor.
- 1) 6.b.1) 2) WHEN Reactor is verified tripped, THEN stop affected RCP(s).
- 2) 6.b.2) 3) Initiate 19000-C, E-0 REACTOR TRIP OR
SAFETY INJECTION.
- 3) 6.b.3) S Approved By M.G. Brill Vogtle Electric Generating Plant Procedure Version 13003-1 47.1 Effective Date 06/12/2013 REACTOR COOLANT PUMP OPERATION Page Number 7 of 42 INITIALS Printed October 2, 2013 at 13:30 2.2.8 The following starting duty cycle for the RCP should be observed: ________
Only one RCP shall be started at any one time.
Two successive starts are permitted, provided the motor is permitted to coast to a stop between starts.
A third start may be made when the winding and core have cooled by running for a period of 20 minutes, or by standing idle for a period of 45
minutes.
2.2.9 During
RCS filling and venting, RCS pressure must be greater than 325 psig prior to starting an RCP to verify adequate seal D/P is
maintained throughout RCS fill and vent. If necessary, the RCP
should be stopped prior to seal D/P dropping less than 200 psid. If
the seal D/P goes below 200 psid during pump operation or coast down, the RCP should be evaluated before restarting the RCP. ________
2.2.10 An RCP shall be stopped IF any of the following conditions exist: ________
Motor bearing temperature exceeds 195°F.
Motor stator winding temperature exceeds 311°F.
Seal water inlet temperature exceeds 230°F Total loss of ACCW for a duration of 10 minutes.
RCP shaft vibration of 20 mils or greater.
RCP frame vibration of 5 mils or greater.
Differential pressure across the number 1 seal of less than 200 psid.
2.2.11 If a loss of RCP seal cooling (Seal Injection and/or ACCW to Thermal barrier) occurs, resulting in RCP shutdown due to exceeding operating limits, then the unit should be cool ed down to Mode 5 to facilitate recovery. Upon reaching Mode 5, ACCW to the Thermal barrier
should be restored. Seal injection should then be returned to service.
This sequence should prevent seal damage, RCP shaft bowing, ACCW System damage, etc. due to excessive thermal stresses. ________
- 1. 028AG2.4.8 001/LOIT AND LOCT/SRO/C/A 3.8/4.5/028AG2.4.8/LO-TA-37021///
Initial condition: - Unit 1 reactor trip and SI occurred.
Current conditions:
- 19011-C, "SI Termination," Step 21, is in progress to evaluate if a bubble exists in the pressurizer.
- Controlling pressurize r level channel, 1LT-459, fails low.
- Actual pressurizer level is 92% and slowly rising.
Which one of the following completes the following statement?
Per 10020-C, "EOP and AOP Rules of Usage," the Shift Supervisor __(1)__
direct the use of 18001-C, "Systems Instrumentati on Malfunction," guidance to restore pressurizer heaters to service while performing 19011-C actions, and in response to actual pressurizer level, the Shift Supervisor
__(2)__ required to transition to 19261-C, "Response to High Pressurizer Level." __(1)__ __(2)__ may is NOT may is may NOT is NOT may NOT is A.B.C.D.K/A028 Pressurizer Le vel MalfunctionG2.4.8 Knowledge of how abnormal operating procedures are used inconjunction with EOPs.K/A MATCH ANALYSIS The question is requires the candidate to determine if AOP 18001-C can be used in conjunction with EOP 19011-C to mitigate complications arising from a failed pressurizer level inst rument. The question is elevated to the SRO level by requiring the candidate to make decision on whether to direct implementation of procedures duringEOP implementation. Friday, February 21, 2014 3:33:47 PM 1
EXPLANATION OF REQUIRED KNOWLEDGE Per EOP and AOP rules of usage 10020-C step 3.
5.9, other procedures such as AOPs and ARPs may be performed in parallel with EOPs as long as their actions do not conflict with the EOP steps. EOP actions take priority.
In both steps 21 and 26 of EOP 19011-C, pressurizer heaters are required to be energized to saturate the pressurizer. With pressurizer level at 92% and rising, th is becomes a crucial action, however the step cannot be accomplished due to heaters being tripped o ff as a result of LT-459 failing low. The use of 18001-C to select away from the failed channel and restore pressurizer heaters is a prudent and necessary action.With pressurizer level at 92% and rising, a yellow path on the CSFST for INVENTORY should exist. Per 1 9200-C step 1 RNO, if a yellow path exist, then initiate FRP based on plant conditions with Shift Supervisor approval.
Per step 11 RNO, transition to FRP 19261-C should only be made if solid plant condtions exist, which are not currently present. Therefore, 19261-C is not required to be entered and should not be entered at this time. Instead, the prio rity should be placed on restor ing pressurizer heaters andcontinuing efforts to lower pressurize r level via charging/letdown mismatch.ANSWER / DISTRACTOR ANALYSISA. Correct. Part 1 is correct.
Per 10020-C 'EOP and AOP Rules of Usage', Other procedures such as AOPs or ARPs may be
performed in parallel with EOPs as long as their actions do
not conflict with the EOP steps. EOP actions take priority.
Additionally, performance of 18001-C to restore pressurizer heaters is prudent and necessary.Part 2 is correct. 19200-C 'C ritical Safety Function Status Trees', states IF a Yellow condition exists, THEN initiate FRP after evaluating plant conditi ons with Shift Supervisor's approval. 19011-C "SI Terminat ion' states IF solid plant conditions are present, THEN refer to 19261 C, FR I.1 Response to High Pressurizer Level. Solid plant conditions are not present. Entry into 19261-C would actually impede the success path in 19011-C.B. Incorrect. Plausible. Part 1 is correct. See Part 1 of choice A above.Part 2 is incorrect but 'plausible' because the candidatesknow that EOP actions always take priority. The
candidate may not realize 192 61-C does not contain any steps that would be differ ent from those already in progress in 19011-C and believ e transition to 19261-C is necessary to prevent going solid in the pressurizer.
C. Incorrect. Plausible. Part 1 is inco rrect but plausable because 10020-C 'EOP andAOP Rules of Usage', Other procedures such as AOPs or ARPs may be performed in parallel with EOPs
as long as their actions do not conflict with the EOP steps.
EOP actions take priority. Friday, February 21, 2014 3:33:47 PM 2
The candidate may not realize t hat the sucess path for current plant conditions resides in re storing pressuir zer heaters and deems use of 18001-C as unnece ssarily slowing the progress of the EOP 19011-C.
Part 2 is correct. See Part 2 of choice A above.
D. Incorrect. Plausible. Part 1 is in correct. See Part 1 of choice C above.
Part 2 is incorrect. See Part 2 of choice B above.SRO JUSTIFICATION(5) Assessment of facility conditions and selection of appropriate proceduresduring normal, abnormal, and emergency situations.-Can the question be answered solely by knowing "systems knowledge", i.e.,how the system works, flowpath, logic, component location?
No, the questionrevolves around transition decisions and does not involve system knowledge.-Can the question be answered solely by knowing immediate operator actions?
No, the question does not address any IOAs.-Can the question be answered solely by knowing entry conditions for AOPs orplant parameters that require direct entry to major EOPs? No, specificknowledge of procedure steps contained within the EOP and AOPs is needed inaddition to entry conditions.
-Can the question be answered solely by knowing the purpose, overall sequenceof events, or overall mitigative strategy of a procedure?
No, specific knowledgeof procedure steps contained within the EOP and AOPs is need ed in addition tooverall knowledge of th e associated procedures.-Does the question re quire one or more of the following?* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or se ction of a procedure to mi tigate, recover, or with which to proceed* Knowledge of when to impl ement attachments and appendices, including how to coordinate these it ems with procedure steps* Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific subprocedures or emergencycontingency procedures Yes, the question requires specific knowledgeof the RNO of step 21 as well as the specifics of both 18001-C and19261-C to determine if tr ansition to these procedures would mitigate thechallenges currently observed.* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant norma l, abnormal, and emergency proceduresFriday, February 21, 2014 3:33:47 PM 3
Level: SROTier # / Group # T1 / G2 K/A# 028G2.4.8 Importance Rating: 3.8 / 4.5Technical
Reference:
EO P 10020-C, Rev 9.0, page 10 EOP 19011-C, Rev 29.2, page 12 & 18 EOP 19200-C, Rev 24.2, page 2, 3, & 10References provided: None Learning Objective: LO-TA-37021 Res pond to High Pre ssurizer Level per 19261-CLO-TA-60030 Respond to a Failu re of Pressurizer Level Instrumentation per 18001-CLO-TA-37005 Terminate Safety Injection per 19000-C or 19011-CLO-TA-05003 Respond to CSFST Trouble alarm and evaluate CSFSTs using SPDS and the
PSMS per 17006-1/2, 13521-1/2, 13505-1/2, and 19200-C LO-LP-60301-12 Given that t he pressurizer level control selector switch is in the NORMAL position
(459/460), describe how and why the plant
will respond to the following instrument
failures. Consider each separately and
include effects on pre ssurizer level control, alarms, RPS, and ESF actuations.
- b. 459 fails low LO-LP-37002-09 Using EOP 1 9200, as a guid e, briefly describe how the steps are accomplished.Question origin: NEW Cognitive Level: C/A 10 CFR Part 55 Content: 43.5 Comments: You have completed the test!Friday, February 21, 2014 3:33:47 PM 4
Approved By Vogtle Electric Generating Plant Procedure Version J. B. Stanley 19011-C 29.2 Effective Date ES-1.1 SI TERMINATION Page Number 05/01/2013 12 of 32 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED Printed September 27, 2013 at 12:07
- 20. Align CCP suction to VCT:
- 20. 20 a. Open VCT OUTLET ISOLATION valves: a. 20.a LV-0112B LV-0112C
- b. 20.b LV-0112D LV-0112E
- 21. Control PRZR pressure:
- 21. 21 a. Check Stub Busses -
ENERGIZED:
- a. Energize Stub Busses by performing the following as
necessary:
21.a NB01 NB10 NB01 NB10
- 1) Open breaker NB01-01 2) Close breaker AA02-22 3) Close breaker NB01-01 1) Open breaker NB10-01 2) Close breaker BA03-18 3) Close breaker NB10-01 b. Check for a bubble in the PRZR to enhance RCS pressure
control. b. Control charging and letdown flows to avoid
sudden pressure changes.
21.b Energize PRZR Heaters.
IF solid plant conditions are present, THEN refer to 19261-C, FR-I.1 RESPONSE TO
HIGH PRESSURIZER
LEVEL. Go to Step 22.
Step 21 continued on next page
Approved By Vogtle Electric Generating Plant Procedure Version J. B. Stanley 19011-C 29.2 Effective Date ES-1.1 SI TERMINATION Page Number 05/01/2013 18 of 32 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED Printed September 27, 2013 at 12:09
- 26. Check RCP status:
- 26. 26 a. RCPs - ALL STOPPED.
- a. Go to Step 27.
26.a b. Check RVLIS full range indication - GREATER THAN 94%.
- b. Perform the following:
26.b Raise PRZR level greater than 90% [90%
ADVERSE].
Raise RCS Subcooling based on core exit TCs greater than 60°F [74°F
ADVERSE].
Use PRZR Heaters, as necessary to saturate
the Pressurizer water.
RCP 1 preferred)
- c. IF an RCP can NOT be started, THEN verify natural circulation using ATTACHMENT B.
26.c IF natural circulation NOT established, THEN raise rate of dumping steam using Steam Dumps.
After natural circulation is verified, maintain rate of
dumping steam.
S Approved By Vogtle Electric Generating Plant Procedure Version J. B. Stanley 19200-C 24.2 Effective Date F-0 CRITICAL SAFETY FUNCTION STATUS TREES Page Number 7/25/12 2 of 11 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED Printed February 21, 2014 at 14:30 NOTES If SPDS display of the Plant Computer is not operable or questionable, manual monitoring of CSFSTs should be performed by a licensed operator. CSFSTs should be monitored continuously if a RED or ORANGE condition is present or each 10 to 15 minutes if the highest priority CSFST is no higher than YELLOW. CSFSTs should be checked in order listed. Priority of operator action is given by the following: Red (Solid) Path - Extreme cha llenge, in Tree Order per Step 1. Orange (Dashed) Path - Severe chall enge, in the Tree Or der per Step 1. Yellow (Dotted) Path - Not sati sfied, in Tree Or der per Step 1. Green (Outlined) Path - Satisfied. If using the Plant Computer (if available) to monitor CSFSTs: The mode indication of the Plant Com puter CSFSTs should be indicating zero. RCP breakers should be opened for RCPs NOT ru nning in order to provide proper RVLIS indication. If SPDS is operable, CSFSTs may be checked by scanning the display console for alarm conditions. Color status of CSFSTs will also be indicate d by letter R for red, O for orange, Y for yellow, G for green, and M for magenta. CSFSTs will indicate active (alarming) pat hs as solid lines and non-active paths as empty or hollow lines.
- 1. Check CSFSTs- SATISFIED:
- 1. IF a Red condition exists, THEN immediately go to FRP.
1 a. Subcriticality (F-0.1)
IF an Orange condition exists, THEN go to FRP after completion of present pass thru CSFSTs.
1.a b. Core Cooling (F-0.2) 1.b Step 1 continued on next page
Approved By Vogtle Electric Generating Plant Procedure Version J. B. Stanley 19200-C 24.2 Effective Date F-0 CRITICAL SAFETY FUNCTION STATUS TREES Page Number 7/25/12 3 of 11 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED Printed February 21, 2014 at 14:30
- c. Heat Sink (F-0.3)
IF a Yellow condition exists, THEN initiate FRP after evaluating plant conditions with Shift Supervisor's approval.
1.c d. Integrity (F-0.4) 1.d e. Containment (F-0.5) 1.e f. Inventory (F-0.6) 1.f 2. Report change of status of any CSFST to the Shift Supervisor, if necessary (i.e., change in status not understood).
- 2. 2 3. Check EOP usage - NO LONGER REQUIRED.
- 3. Return to Step 1.
3 4. Monitoring of CSFSTs is no longer required.
4 END OF PROCEDURE TEXT
Approved By Vogtle Electric Generating Plant Procedure Version J. B. Stanley 19200-C 24.2 Effective Date F-0 CRITICAL SAFETY FUNCTION STATUS TREES Page Number 7/25/12 10 of 11 Sheet 1 of 1 Printed February 21, 2014 at 14:30 RVLIS DYNAMIC RANGE:> 95% -4 RCP> 68% -3 RCP
> 46% -2 RCP> 35% -1 RCPRVLIS INDICATES UPPER HEAD GREATER THAN 98% FULLPRESSURIZER LEVEL AT LEAST ONE RCP RUNNINGPRESSURIZER LEVEL F-0.6 INVENTORYAT LEAST ONE RCP RUNNINGRVLIS INDICATES UPPER HEAD GREATER THAN 98% FULLRVLIS DYNAMIC RANGE:> 95% -4 RCP> 68% -3 RCP> 46% -2 RCP> 35% -1 RCPLESS THAN 92%
Approved By C.S. WALDRUP Vogtle Electric Generating Plant Procedure Number Rev 10020-C 9 Date Approved 01/26/2011 EOP AND AOP RULES OF USAGE Page Number 10 of 27 Printed September 27, 2013 at 12:05 3.5.8 ES-0.0, REDIAG NOSIS, may be entered any time based on operator judgement and may be entered as follows:
3.5.8.1 ES-0.0 may be enter ed when there is doubt in being in correct EOP.
3.5.8.2 Safety injection is in service or is required.
3.5.8.3 E-O, REACTOR TRIP OR SAFETY INJECTION, has been executed and a transition has been made to another EOP which is bounded by the ORGs only (not FRGs).
3.5.9 Other
procedures such as AOPs or ARPs may be performed in parallel with EOPs as long as their actions do not c onflict with the EOP steps. EOP actions take priority.
3.6 STEP PLACE-KEEPING 3.6.1 When exiting an EOP step it is necessa ry to track what procedure and step was exited such that when directed to "return to procedure step in affect", the correct procedure step may be re-entered. A red ribbon page marker has been provided in the Simulator and Main Control Room EO P sets for assistance in tracking such transitions.
3.6.2 Step by step place-keeping is a va luable human performance tool. It shall be performed in accordance with plant standard and management expectations.
3.7 NOTES
AND CAUTIONS All NOTES and CAUTIONS shall be reviewed by the Shift Supervisor. Those NOTES or CAUTIONS that are pertinent to the evolution in progress shall be read aloud to the operating crew.
3.8 MODES
OF APPLICABILITY 19000-C E-0 1,2,3 Assumes RHR system not in service and SI operable
19001-C ES-01 1,2 Assumes trip from power
19002-C ES-0.2 1,2.3 Assumes No-load conditions
19003-C ES-0.3 1,2,3 Assumes No-load conditions
19004-C ES-0.4 1,2,3 Assumes No-load conditions
- 1. 032AA2.03 001/LOIT AND LOCT/SRO/C/A 2.8/3.1/032AA2.03/LO-LP-39213-04///
Given the following conditions:
- Unit 1 is in Mode 6.
- Source Range N31/32 each indicate ~10 cpm.
- Source Range N31 is powered from 1AY1A.
- Source Range N32 is powered from 1NLP39.
Which one of the following completes the following statement?
Tech Spec LCO 3.9.3, "Nuclear Instrument ation," __(1)__ met, If the Source Range N32 detector high volt age power supply breaker were to trip on overcurrent, the OATC woul d observe the QMCB N32 meter indicating __(2)__. __(1)__ __(2)__ is as-is is bottom of scale is NOT as-is is NOT bottom of scale A.B.C.D.K/A 032 Loss of Source Range NI AA2.03 Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation: - Expected values of source range indication when high voltage isautomatically removedK/A MATCH ANALYSIS The question tests the candidate's ability to determine the indi cation of the Source Range NIs following the trip of the high volt age power supply breaker in Mode 6 with 10CPM indicated.(Note: At Vogtle, Source Range NIs high voltage power supp lies are no longer automatically removed - GAMAMETRICS now installed.)EXPLANATION OF REQUIRED KNOWLEDGEWednesday, February 26, 2014 8:20:39 AM 1
The candidate is required to know TS 3.9.3 Bases to determine t he Operability of the Source Range NIs. In M ode 6, one SR NI can be powe red from a Non-1E power supply provided the other SR NI is powered from it's normal 1E power supply.
Additionally, the candidate is required to determine the SR NI indication following a loss of the high voltage power supply. In this situation, the NI det ector would be down powered and the indication would drop to bottom of scale.ANSWER / DISTRACTOR ANALYSISA. Incorrect. Plausible. The first part is correct. When any of the safety-related busses supplying power to one of the detectors (NI-0031 or NI-0032) associated with the source r ange neutron flux monitors are taken out of service, the co rresponding source range neutron flux monitor may be considered OPERABLE when its detector is powered from a temporary nons afety-related power source, provided the detector for the oppo site source range neutron flux monitor is powered from its normal source.
The second part is incorrect.
Loss of the hi gh voltage power supply will constitute a complete loss of power to the SR NI detectors, resulting in indication failing to bottom of scale.
However, the NI's have 3 diff erent power supplies that support various functions.
If a loss of control pow er only occurs, all bistables trip, however NI in dication is unchanged and remains"as-is". Therefore, this distractor is plausible.B. Correct. The first part is correct. See the first part of choice A above.
The second part is corre ct. The loss of the High Voltage Power Supply will result in a complete loss of pow er to the SR NIs and indication will read bottom of scale.
C. Incorrect. Plausible. The first part is in correct. Per TS 3.9.3 Bases, one SR NI can be powered from a Non-1E power supply provided the other SR NI is powered from it's normal 1E power supply in Mode 6. It is abnormal for a Safety-Related, Tech Spec required component to be powered from a non-1E power supply and be considered OPERABLE. It is reasonabl e for a candidate without knowledge of this s pecific exception to determine the LCO not met. Therefore, this distractor is plausible.
The second part is inco rrect. See the second part of choice A above.D. Incorrect. Plausible. The first part is incorrect. See the first part of choice C above.
The second part is correct. See the second part of choice B above.SRO JUSTIFICATION (10CFR43(b))Wednesday, February 26, 2014 8:20:39 AM 2
(2) Facility operating limitations in the technical specifications and their bases.-Can question be answered solely by knowing = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action?
No, theknowledge required is not included in any TS or TRM action.-Can question be answered solely by knowing the LCO/TRM information listed "above-the-line?"
No, the knowledge required do es not exist above the line inany TS or TRM.-Can question be answered solely by knowing the TS Safety Limits?
No, SR NIs are not discussed in the TS Safety Limits.-Does the question involve one or more of the following for TS,TRM, or ODCM?
- Application of Required Actions (Sec tion 3) and Surve illance Requirements (Section 4) in accordance with rules of applicati on requirements (Section 1)* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)* Knowledge of TS bases that is required to analyze TS required actionsand terminology.
Yes, specific knowledge of the Mode 6 power supply alignments listed in TS Bases 3.9.3 are required to determine if the LCOis met.Wednesday, February 26, 2014 8:20:39 AM 3
Level: SROTier # / Group # T1 / G2 K/A# 032AA2.03 Importance Rating: 2.8 / 3.1Technical
Reference:
17010-1 Rev 50 , page 60Tech Spec Bases 3.9.3 Rev 3-4/09, page B3.9.3-1&2References provided: None Learning Objective: LO-LP-60302-05 Describe how and why a reactor startup would be affected by a source range
instrument failure wh en the reactor is at the following power levels: above and
below P-6.
LO-PP-17201-01 Discuss the operation of the Source &
Intermediate Range De tectors to include:
- a. Type of detector
- b. Gamma compensation
- c. When they are used
- g. Power supplies (also including theeffects on loss of inst rument or control power)LO-LP-39213-04 Descr ibe the bases for any given Tech Spec in section 3.9.Question origin: NEW Cognitive Level: C/A 10 CFR Part 55 Content: 41.7 / 43.2 Comments: You have completed the test!Wednesday, February 26, 2014 8:20:39 AM 4
Nuclear Instrumentation B 3.9.3 Vogtle Units 1 and 2 B 3.9.3-1 Rev. 3-4/09 B 3.9 REFUELING OPERATIONS
B 3.9.3 Nuclear Instrumentation
BASES BACKGROUND The source range neutron flux monitors are used during refueling operations to monitor the core reactivity condition. The installed
source range neutron flux monitors (NI-0031 and NI-0032) are part of the Nuclear Instrumentation System (NIS). These detectors are located external to the reactor vessel and detect neutrons leaking from the core. Temporary neutron flux detectors which provide equivalent indication may be utilized in place of installed
instrumentation.
The installed source range neutron flux monitors are fission chamber detectors. The detectors monitor the neutron flux in counts per second. The instrument range covers seven decades of neutron flux (1E-1 cps to 1E +6 cps) with a 2% instrument accuracy. The
detectors also provide continuous visual indication in the control room.
The NIS is designed in accordance with the criteria presented in Reference 1.
APPLICABLE Two OPERABLE source range neutron flux monitors are required SAFETY ANALYSES to provide a signal to alert the operator to unexpected changes in core reactivity such as an improperly loaded fuel assembly. The need for a safety analysis for an uncontrolled boron dilution accident is minimized by isolating all unborated water sources except as provided for by LCO 3.9.2, "Unborated Water Source Isolation Valves."
The source range neutron flux monitors satisfy Criterion 3 of 10 CFR 50.36 (c)(2)(ii).
LCO This LCO requires that two source range neutron flux monitors be OPERABLE to ensure that redundant monitoring capability is available to detect changes in core reactivity. To be OPERABLE each monitor must provide visual indication.
When any of the safety-related busses supplying power to one of the detectors (NI-0031 or NI-0032) associated with the source range neutron flux monitors are taken out of service, the corresponding source range neutron flux monitor may be considered OPERABLE when its detector is powered from a temporary nonsafety-related (continued)
Nuclear Instrumentation B 3.9.3 Vogtle Units 1 and 2 B 3.9.3-2 Rev. 1-4/09 BASES LCO source of power, provided the detector for the opposite source range (continued) neutron flux monitor is powered from its normal source.
APPLICABILITY In MODE 6, the source range neutron flux monitors must be OPERABLE to determine changes in core reactivity. There are no other direct means available to check core reactivity levels. In MODES 2, 3, 4, and 5, the operability requirements for the installed source range detectors and circuitry are specified in LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation." ACTIONS A.1 and A.2 With only one source range neutron flux monitor OPERABLE, redundancy has been lost. Since these instruments are the only
direct means of monitoring core reactivity conditions, CORE ALTERATIONS and positive reactivity additions must be suspended immediately. Performance of Required Action A.1 shall not preclude completion of movement of a component to a safe position or normal cooldown of the coolant volume for the purpose of system temperature control.
B.1 Condition B is modified by a Note to clarify the requirement that entry into or continued operation in accordance with Condition A is required
for any entry into Condition B. The Note reinforces conventions of LCO applicability as stated in LCO 3.0.2 and as reflected in examples in 1.3, Completion Times.
With no source range neutron flux monitor OPERABLE, action to restore a monitor to OPERABLE status shall be initiated immediately.
Once initiated, actions shall be continued until a source range neutron flux monitor is restored to OPERABLE status.
B.2 With no source range neutron flux monitor OPERABLE, there are no direct means of detecting changes in core reactivity. However, since CORE ALTERATIONS and positive reactivity additions are not to be (continued)
Approved By J.B. Stanley Vogtle Electric Generating Plant Procedure Number Rev 17010-1 50 Date Approved 08/16/2011 ANNUNCIATOR RESPONSE PROCE DURES FOR ALB 10 ON PANEL 1C1 ON MCB Page Number 3 of 66 Printed September 27, 2013 at 13:52 ALB 10 (1) (2) (3) (4) (5) (6)
A SR/IR SIG PROCESSOR
TROUBLE
NIS SOURCE AND INTMD RANGE TRIP BYPASS
POWER RANGE HI NEUTRON FLX HI
SETPOINT ALERT
REACTOR BYPASS BRKR BYA IN-OPERATE
REACTOR BYPASS BRKR BYA CLOSE
ROD CONTROL
NON URGENT FAILURE B
SOURCE RNG HI SHUTDOWN FLUX
ALARM BLOCKED
POWER RANGE HI NEUTRON FLX LOW SETPOINT
REACTOR BYPASS BRKR BYB IN-OPERATE
REACTOR BYPASS BRKR BYB CLOSE ROD CONTROL URGENT FAILURE
C SOURCE RANGE HI FLUX LEVEL AT SHUTDOWN
POWER RANGE
CHANNEL DEVIATION OVERPOWER T ROD BLOCK AND
RUNBACK ALERT
ROD BANK LO LIMIT
RPI NON URGENT
ALARM NIS CHANNEL
ON TEST
D INTMD RANGE HI FLUX LEVEL ROD STOP
PWR RANGE UP DET HI FLX DEV
OVERPOWER
ROD STOP ROD BANK LO-LO LIMIT
RPI URGENT ALARM
ROD DEV E
SR/IR REMOTE
SIG PROCESSOR
DPU-B TROUBLE
OVERTEMP T ROD BLOCK AND
RUNBACK ALERT
ROD AT BOTTOM
RADIAL TILT
F SR/IR AMPLIFIER
TROUBLE
POWER RANGE HI NEUTRON FLX
RATE ALERT
ROD DRIVE M-G
SET TROUBLE
TWO OR MORE RODS AT BOTTOM
DELTA FLUX
DEVIATION
Approved By J.B. Stanley Vogtle Electric Generating Plant Procedure Number Rev 17010-1 50 Date Approved 08/16/2011 ANNUNCIATOR RESPONSE PROCE DURES FOR ALB 10 ON PANEL 1C1 ON MCB Page Number 60 of 66 Printed September 27, 2013 at 13:52 WINDOW F01 ORIGIN SETPOINT NC-35M NC-36M
Not Applicable
SR/IR AMPLIFIER
TROUBLE
1.0 PROBABLE
CAUSE
- 1. High Voltage Power Supply great er than 875V or less than 660V.
- 2. Loss or degraded +15v Power Supply in WR amplifier (or isolator for N36)
- 3. Loss or degraded -15v Power Supply in WR amplifier (or isolator for N36)
- 4. Degraded +5V Power Supply(s) in Isolator Assembly (N36 only)
2.0 AUTOMATIC
ACTIONS NONE
3.0 INITIAL
OPERATOR ACTIONS Go to 18002-C, "Nuclear Instrument ation System Malfunction".
4.0 SUBSEQUENT
OPERATOR ACTIONS NONE
5.0 COMPENSATORY
OPERATOR ACTIONS NONE
END OF SUB-PROCEDURE
REFERENCE:
1X6AS01-154
- 1. 033A2.03 001/LOIT AND LOCT/SRO/C/A 3.1/3.5/033A2.03/LO-TA-25010///
Procedure 13719-1, "Spent Fuel Pool Cooling and Purificati on," sections as follows:
- Section 4.2.2, "SFP Makeup from the RWST thro ugh the SFP Purification Loop" - Section 4.2.4, "SFP Makeup from the RMWST" Initial conditions:
- Unit 1 is defueled.
- Transfer canal is drai ned for transfer cart inspection. - Spent fuel shuffle is in progress in the FHB.
Current conditions:
- ALB05-E02 SPENT FU EL PIT LO LEVEL is received.
- Personnel in the FHB re port SFP level is slowly lowering.
- 18030-C, "Loss of Spent F uel Pool Level or Cooling," is entered.
Which one of the following completes the following statement?
To mitigate the consequences of the event, t he Shift Supervisor is required to direct makeup to the SFP usi ng 13719-1, Section __(1)__, and per the Bases of Tech Spec 3.7.15, "Fuel Storage Pool Water Level," maintaining the required minimum water level in the SF P __(2)__ ensur e adequate iodine decontamination factors are met for a fuel handling accident. __(1)__ __(2)__ 4.2.2 does 4.2.2 does NOT 4.2.4 does 4.2.4 does NOT A.B.C.D.K/A 033 Spent Fuel Pool Cooling A2.03 Ability to (a) predict the impacts of the following malfunctions oroperations on the Spent Fuel Pool Cooling System; and (b) based onthose predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Abnormal spentfuel pool water level or loss of water levelFriday, February 21, 2014 3:41:24 PM 1
K/A MATCH ANALYSIS The question tests the candidate's ability to predict the impact of low spent fuel pool level by having to recall the bases for TS 3.7.
15 level. The candidate is also required to mitigate the consequence of the event through the selection of a makeup source by selecting the appropriate pr ocedure section to perform.EXPLANATION OF REQUIRED KNOWLEDGE Per ARP 17005-1, ALB05-E02 alarms at a Spent Fuel Pool Lev el of 217'-0". Reports from the field have verified and low and decr easing pool level. AOP 18030-C is entered to mitigate the event. Per step 6 of this procedure, makeup to the SFP per SOP 13719-1 is directed. The specific procedure section is not specif ied. The candidate must recognize that makeup due to leakage will be from the RWST and not the RMWST. This ensures SFP boron concentrat ion will be maintained. This requirement is stipulated in SOP 13719-1 Precaution and Limitation 2.1.7 and in a CAUTION at the beginning of sections 4.2.3 and
4.2.4. These
state that non-borated makeup is usually only allowed for normal evaporative level losses. Borated water sources are the preferred maekup source fo r abnormal or unexplained level lo sses. Objective LO-PP-25102-11 is utilized during LOIT to re-enforce the us e of borated water sources only during leakage.With SFP level less than the Tech Spec limit of 217'-0", less than 23 feet of water existover the spent fuel stored in the racks. Per TS 3.7.15 Bases, the minimum water level in the fuel storage pool meets the assump tions of iodine decon tamination factors following a fuel handling acci dent. The bases also discusse s TS 3.7.15 water level as providing shielding to mini mize general area dose and pr ovide shielding during spent fuel movement.ANSWER / DISTRACTOR ANALYSISA. Correct. The first part is corre ct. Per SOP 13719-1 'Spent Fuel Pool Cooling and Purification System
', makeup due to leakage is from a borated source.
The second part is correct. Per Technical Specification 3.7.15
'Fuel Storage Pool Water Level' bases, the minimum water level in the fuel storage pool meet s the assumptions of iodine decontamination factors followi ng a fuel handling accident.
B. Incorrect. Plausible. Th e first part is correc
- t. See the first part of choice A above.
The second part is inco rrect. Per Technical Specification 3.7.15
'Fuel Storage Pool Water Level' bases, the minimum water level in the fuel storage pool meet s the assumptions of iodine decontamination factors following a fuel handling accident. However, the bases also discusses the shielding function of the water, which is a more commo nly known benefi
- t. A candidate without adequate knowledge of t he TS Bases, may conclude that iodine decontamination is not a factor at all for water level,Friday, February 21, 2014 3:41:24 PM 2
or may assume the normal mi nimum water level of 218'-0" required. Therefore, this distractor is plausible.
C. Incorrect. Plausible. The first part is incorrect. Per SOP 1 3719-1 'Spent Fuel Pool Cooling and Purification System
', makeup due to leakage is from a borated source. However, normally SFP makeup is made from either Demin Water or the RMWST since the level change is due to evaporative loss and the boron is left in solution. A candidate with inadequate knowledge of the
purpose behind using the differ ent sources would find it reasonable to makeup using a norma l source. Therefore, this distractor is plausible.
The second part is correct. See the second part of choice A above.D. Incorrect. Plausible. The first part is in correct. See the first part of choice C above.
The second part is inco rrect. See the second part of choice B above.SRO JUSTIFICATION (10CFR43(b))(2) Facility operating limitations in the technical specifications and their bases.-Can question be answered solely by knowing = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action?
No, the question requires specific know ledge of the bases for TS 3.7.15. The immediate action of the associated RAS does not address the purpose of the level or howto restore it.-Can question be answered solely by knowing the LCO/TRM information listed "above-the-line?"
The information above the line deals with the required levelto be maintained only. It does not address the reason for the level.-Can question be answered solely by knowing the TS Safety Limits?
No, SpentFuel Pool level is not a Safety Limit.-Does the question involve one or more of the following for TS,TRM, or ODCM?
- Application of Required Actions (Sec tion 3) and Surve illance Requirements (Section 4) in accordance with rules of applicati on requirements (Section 1)
- Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)* Knowledge of TS bases that is required to analyze TS required actionsand terminology.
Yes, the reason SFP level is maintain ed >23 ft abovethe fuel is only specified in the Bases for TS 3.7.15.Friday, February 21, 2014 3:41:24 PM 3
Level: SROTier # / Group # T2 / G2 K/A# 033A2.03 Importance Rating: 3.1 / 3.5Technical
Reference:
SOP 13719-1 Rev 55.2, pages 6, 19, & 20 ARP 17005-1 Rev 34.
2, pages 43-45 Tech Spec 3.7.15 Amendm ent No. 158, page 3.7.15-1 Tech Spec Bases 3.7.15 Re v 1-10/01, page B 3.7.5.15-1References provided: None Learning Objective: LO-PP-25102-12 De scribe the minimum allowable water level over spent fuel and the basis for this
level.LO-PP-25102-11 Describe w hen the different sources of makeup to the spent fuel pool would be
used. For evaporation, For leakageLO-TA-25010 Makeup to t he SFP per 13719-1/2, 13903-C, and 18030-C Attachment CQuestion origin: NEW Cognitive Level: C/A
10 CFR Part 55 Content: 41.10 / 43.2 Comments:
Early submittal 401-9 response:-Need to make sure Secti on 4.2.4 cannot be argued as a correct answer. 13719 and EOP caution say borated water "should" rather than "shall" be used. -Having Keff requirements as a basis for the SFP waterlevel does not seem plausible. However, I think this
can be solved by reframing the second question to
state "per Bases of Tech Spec 3.7.15", maintain the
required minimum wa ter level in the SFP does/does not ensure adequate idodine decont amination factors are met.- JAT 12/19/13 (Editorial)
The new question incorporat es the above suggestion, and the first concern with the orig inal question has a learning objective to reinforce the use of borated water sources.
Need to make sure this is definitive enough as a "technicalsource."
- JAT 2/4/2014You have completed the test!Friday, February 21, 2014 3:41:24 PM 4
Fuel Storage Pool Water Level 3.7.15 Vogtle Units 1 and 2 3.7.15-1 Amendment No. 158 (Unit 1) Amendment No. 140 (Unit 2) 3.7 PLANT SYSTEMS 3.7.15 Fuel Storage Pool Water Level
LCO 3.7.15 The fuel storage pool water level shall be 23 ft over the top of irradiated fuel assemblies seated in the storage racks.
APPLICABILITY: During movement of irradiated fuel assemblies in the fuel storage pool.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Fuel storage pool water level not within limit.
A.1 -------------NOTE--------------
LCO 3.0.3 is not applicable.
Suspend movement of irradiated fuel assemblies in the fuel storage pool.
Immediately
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.15.1 Verify the fuel storage pool water level is 23 ft above the top of the irradiated fuel assemblies seated in the storage racks.
In accordance with the Surveillance Frequency Control Program
Fuel Storage Pool Water Level B 3.7.15 Vogtle Units 1 and 2 B 3.7.15-1 Rev. 1-10/01 B 3.7 PLANT SYSTEMS B 3.7.15 Fuel Storage Pool Water Level BASES BACKGROUND The minimum water level in the fuel storage pool meets the assumptions of iodine decontamination factors following a fuel handling accident. The specified water level shields and minimizes the general area dose when the storage racks are filled to their maximum capacity. The water also provides shielding during the movement of spent fuel. A general description of the fuel storage pool design is given in the FSAR, Subsection 9.1.2 (Ref. 1). A description of the Spent Fuel Pool Cooling and Cleanup System is given in the FSAR, Subsection 9.1.3 (Ref. 2). The assumptions of the fuel handling accident are given in the FSAR, Subsection 15.7.4 (Ref. 3). APPLICABLE The minimum water level in the fuel storage pool meets SAFETY ANALYSES the assumptions of the fuel handling accident described in Regulatory Guide 1.25 (Ref. 4). The resultant 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose per person at the exclusion area boundary is a small fraction of the 10 CFR 100 (Ref. 5) limits. According to Reference 4, there is 23 ft of water between the top of the damaged fuel bundle and the fuel pool surface during a fuel handling accident. With 23 ft of water, the assumptions of Reference 4 can be used directly. In practice, this LCO preserves this assumption for the bulk of the fuel in the storage racks. In the case of a single bundle dropped and lying horizontally on top of the spent fuel racks, however, there may be < 23 ft of water above the top of the fuel bundle and the surface, indicated by the width of the bundle. To offset this small nonconservatism, the analysis assumes that all fuel rods fail, although analysis shows that only the first few rows fail from a hypothetical maximum drop. The fuel storage pool water level satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii). (continued)
Approved By J. B. Stanley Vogtle Electric Generating Plant Procedure Number Rev 13719-1 55.2 Date Approved 03/27/2012 SPENT FUEL POOL COOLING AND PURIFICATION SYSTEM Page Number 6 of 82 INITIALS Printed October 2, 2013 at 14:49
2.0 PRECAUTIONS
AND LIMITATIONS
2.1 PRECAUTIONS
2.1.1 The SFPCPS should be operated as necessary to maintain the SFP temperature below the high te mperature alarm setpoint of 130°F. ________
2.1.2 The differential pressure ac ross the SFP Skimmer Filter should not exceed 20 psid. ________
2.1.3 The differential pressure across the SFP Purification Loop cartridge filter should not exceed 70 psid. ________
2.1.4 The purification flow thr ough the SFP Demineralizer System should not exceed 120 gpm. ________
2.1.5 Thoroughly
fill and vent all applicable SFPCPS components prior to returning them to service a fter maintenance. This minimizes system performance degradation due to gas entrainment. ________
2.1.6 Any time that water is being removed from the RWST for makeup to the SFP or when the Refueling Water Purification
Pump is taking suction from the RWST, the RWST level shall be
maintained above the applicable Technical Specification low limit. ________
2.1.7 Non-borated makeup is usually only allowed for normal evaporative level losses. Borated water sources are the
preferred makeup source for abnormal or unexplained level losses. ________
2.1.8 IF in Modes 1-4 opening of 1-1204-U4-158 must NOT be performed, since this would result in RWST declared inoperable ________
Approved By J. B. Stanley Vogtle Electric Generating Plant Procedure Number Rev 13719-1 55.2 Date Approved 03/27/2012 SPENT FUEL POOL COOLING AND PURIFICATION SYSTEM Page Number 19 of 82 INITIALS Printed October 2, 2013 at 14:49 CAUTION SFP boron concentration should be checked following makeup to assure a minimum boron concentration of 2000 ppm.
4.2.3 SFP Makeup from the Demineralized Water System (SNC16999) (SNC12987)
CAUTION Non-borated makeup is only allowed for normal evaporative level losses.
Abnormal or unexplained level losses should be compensated for by
using only borated sources.
4.2.3.1 Open SFP DEMIN WTR SPLY ISO, 1-1213-U4-055. (RA53) ________
CAUTIONS When gravity filling from the RWST, t he Spent Fuel Pool level must be monitored continuously to pr event overflowing the SFP. Spent fuel Pool lighting rec eptacles are at the 218'9'" level.
Spent fuel Pool HI Level Alarm setpoint is at 219'.
4.2.3.2 Monitor Spent Fuel Pool level (see Figure 1). ________
4.2.3.3 WHEN the required water level is reached, close and lock SFP DEMIN WTR SPLY ISO, 1-1213-U4-055, (RA53); (IV REQUIRED) ________
4.2.3.4 Request Chemistry sample the Spent Fuel Pool to determine the boron concentration. ________
Approved By J. B. Stanley Vogtle Electric Generating Plant Procedure Number Rev 13719-1 55.2 Date Approved 03/27/2012 SPENT FUEL POOL COOLING AND PURIFICATION SYSTEM Page Number 20 of 82 INITIALS Printed October 2, 2013 at 14:49 CAUTION SFP boron concentration should be checked following makeup to assure a minimum boron concentration of 2000 ppm.
4.2.4 SFP Makeup from RMWST (SNC16999) (SNC12987)
CAUTION Non-borated makeup is usually only allowed for normal evaporative level losses. Borated water sources are the preferred makeup source for
abnormal or unexplained level losses.
4.2.4.1 Open SFP CLG RMWST ISOLATION VALVE, 1-1213-U4-054. (RA53) ________
CAUTIONS When gravity filling from the RWST, t he Spent Fuel Pool level must be monitored continuously to pr event overflowing the SFP. Spent fuel Pool lighting rec eptacles are at the 218'9'" level.
Spent fuel Pool HI Level Alarm setpoint is at 219'.
4.2.4.2 Monitor Spent Fuel Pool level (see Figure 1). ________
4.2.4.3 WHEN required water level is reached, close SFP CLG RMWST ISOLATION VALVE, 1-1213-U4-054, (RA53); (IV REQUIRED) ________
4.2.4.4 Request Chemistry sample the Spent Fuel Pool to determine the boron concentration. ________
Approved By C. H. Williams, Jr.
Vogtle Electric Generating Plant Procedure Version 17005-1 34.2 Effective Date 6/21/13 ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 05 ON PANEL 1A2 ON MCB Page Number 43 of 67 Printed December 4, 2013 at 13:27 WINDOW E02 ORIGIN SETPOINT 1-LSHL-625
217 feet elevation
SPENT FUEL PIT LO LEVEL
1.0 PROBABLE
CAUSE
- 1. Insufficient inventory during filling or refueling operation.
- 2. Normal evaporation.
- 3. System leak.
- 4. Loss of air to the Fuel Transfer Canal and/or Cask Loading Pit Gate Seals.
2.0 AUTOMATIC
ACTIONS NONE
3.0 INITIAL
OPERATOR ACTIONS NONE
Approved By C. H. Williams, Jr.
Vogtle Electric Generating Plant Procedure Version 17005-1 34.2 Effective Date 6/21/13 ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 05 ON PANEL 1A2 ON MCB Page Number 44 of 67 Printed December 4, 2013 at 13:27 WINDOW E02 (Continued)
4.0 SUBSEQUENT
OPERATOR ACTIONS
- 1. Dispatch an operator to determine actual level locally. (see Figure 1 in this procedure).
- 2. Notify the Security Alarm Station (C AS) to dispatch a security patrol to check for any indications of sabotage.
- 3. Refer to 13719-1, "Spent Fuel P ool Cooling And Purification" and return the Spent Fuel Pit to normal level (218.5 feet).
- 4. IF level cannot be maintained greater than 217 feet with fuel movement in containment in progress or 216.5 feet with the Spent Fuel Pool Gate Valve closed, THEN suspend movement of irradiated fuel assemblies in the Spent Fuel Pool and all crane operations over the Spent Fuel Pool. Initiate 18030-C, "Loss Of Spent Fuel Pool Level Or Cooling" and 18006-C "Fuel Handling Event.
" 5. Check service air to gate seals and refer to 13710-1, "Service Air System" to restore service air if lost.
- 6. Refer to Technical Specification LCO 3.7.15.
5.0 COMPENSATORY
OPERATOR ACTIONS NOTE If the East and West pools are connected through the cask loading pit, Unit 1 annunciator ALB05E02 will detect a low level condition for both pools. Verify Spent Fuel Pool Level every 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> per 11883-1, "Radwaste Rounds Sheets."
END OF SUB-PROCEDURE
REFERENCES:
1X4DB130, PLS, 1X5DT0037, Technical Specifications LCO 3.7.15 Commitments SNC11369, 1986308950; SNC4521, 1984301472;
SNC16061, 1996332947
Approved By C. H. Williams, Jr.
Vogtle Electric Generating Plant Procedure Version 17005-1 34.2 Effective Date 6/21/13 ANNUNCIATOR RESPONSE PROCEDURES FOR ALB 05 ON PANEL 1A2 ON MCB Page Number 45 of 67 Printed December 4, 2013 at 13:27 Figure 1 - Spent Fuel Pool Local Water Level Indication
LOW ALARM ELEV.219'HIGH ALARM NORMAL LEVEL
- 1) LEVEL NUMBERS ARE PLANT ELEVATIONS IN FEET
- 2) POOL VOLUME APPROX.
453,000 GALS AT "N" 3) 1 FOOT OF POOL (ONLY)
ELEVATION EQUALS APPROX.
11,408 GALS 218' 6" 218'217' 6" 217'
- 1. 055G2.2.42 001/LOIT/SRO/M/F 3.9/4.6/055G2.2.42/LO-TA-60020A///At time 1000:
- Unit 1 is at 100% reactor power.
- 18009-C, "Steam Generator Tube Leak," is in progress. - SG sample results indicate high activity on SG #1.At time 1020:
- 1RE-0724, Steam Li ne Rad Monitor, indicates 105 gpd.
- 1RE-0810, SJAE Exhaust Rad Monito r, indicates 120 gpd.
- 1RE-0724 ROC is 55 gpd/hour.
- 1RE-0810 ROC is 60 gpd/hour.
Which one of the following completes the following statement?
Per Tech Spec 3.4.13, "RCS Operational Leakage," the primary to secondary leakage__(1)__ exceed the limit, and per 18009-C, the Shift Supervisor is required to initiate __(2)__ to lower reac tor power. __(1)__ __(2)__ does 18013-C, "Rapi d Power Reduction" does 12004-C, "Power Operation (Mode 1)" does NOT 18013-C, "R apid Power Reduction" does NOT 12004-C, "Power Operation (Mode 1)" A.B.C.D.K/A 055 Condenser Air RemovalG2.2.42 Ability to recognize system parameters that are entry-levelconditions for Technical Specifications.K/A MATCH ANALYSIS The question requires the candidate to util ize radiation monitors associated with Condenser Air Removal (1RE-0810) to reco gnize entry level fo r Tech Specs on RCS leakage. The candidate is then required to select which proc edure will be utilzied based on the current primary to se condary leakage for plant shutdown.EXPLANATION OF REQUIRED KNOWLEDGEFriday, February 21, 2014 3:43:48 PM 1
Per TS 3.4.13, 150 gpd primary to se condary LEAKAGE through any one steam generator exceeds allo wable RCS operational LEAKAGE. In this condidtion, a shutdown to Mode 3 in 6 hrs and Mode 5 in 36 hrs is required. This shutdown can be accomplished using either t he guidance of 12004-C or 18013-C. The decision is based on the characteristics of the l eak and directed out of 18009-C.
Per 18009-C, if the tube leak is <5gpm and changing at a ra te of < 30 gpd/hr, then a shutdown per 12004-C is suffi ciently aggressive.
If the tube leak is >5gpm or <5 gpm but changing at a rate >30 gp d/hr, then a more agressive shutdown utilizing 18013-C is necessary to ensure the plant is shutdown befor e the leak propogates into a rupture.
The rate of change is de termined using 1RE-0724, N-16 Rad Monitor and/or 1RE-0810, SJAE Exhaust Rad Monitor, whose rate of change indications become valid after 20mintues.This decission tree is a complex set enco mpassing steps 5 thru 10 of 18009-C and is often mis-navigated. A high level understanding of steps are required to ensure internal self-checking of this operational decision.ANSWER / DISTRACTOR ANALYSISA. Incorrect. Plausible. The first part is incorrect. Per TS 3.4.13, 150 gpd LEAKAGE in any steam generator exceeds th e TS limit. Since both rad montiors have leakage below this value, the limit has not be exceeded. However, the thre shold between a tube leak and a tube rupture is 120 gpm.
Therefore, a candidate without sufficient knowledge of the TS lim its could transpose in their minds the 120 and 150 values and conclude that the TS limit has been exceeded. Therefore, this distractor is plausible.
The second part is correct. Pe r 18009-C, with the leak rate <5 gpm and the rate of change >
30gpd/hr, a po wer reduction using 18013-C wo uld be required.
B. Incorrect. Plausible. Th e first part is incorre ct. See the first part of choice A above.
The second part is incorrect.
Per 18009-C, with the leak rate
<5 gpm and the rate of cha nge > 30gpd/hr, a power reduction using 18013-C would be required. However, a candidate without specific knowledge of the procedure decision tree values could conclude that t he leak rate (120 gpd = 0.0833 gpm) is not sufficiently large to justify such an aggressive shutdown as 18013-C. Therefore, this distractor is plausible.C. Correct. The first part is corre ct. Per TS 3.4.
13, 150 gpd LEAKAGE in any steam generator exceeds th e TS limit. Since both rad montiors have leakage below this value, the limit has not be exceeded.The second part is correct. See the second part of choice A above.Friday, February 21, 2014 3:43:48 PM 2
D. Incorrect. Plausible. The first part is correct. The first part is correct. See the first part of choice C above.
The second part is inco rrect. See the second part of choice B above.SRO JUSTIFICATION (10CFR43(b))(5) Assessment of facility conditions and selection of appropriate proceduresduring normal, abnormal, and emergency situations.-Can the question be answered solely by knowing "systems knowledge", i.e.,how the system works, flowpath, logic, component location? No, the procedure direction decision is based out of NEI guidance on leak characteristics. It is not a logic decision, it is purely based on imperical data from the industry.-Can the question be answered solely by knowing immediate operator actions? No, the procedure knowledge required is not an IOA.
-Can the question be answered solely by knowing entry conditions for AOPs orplant parameters that require direct entry to major EOPs?
No, the procedureknowledge required is neither associated with entry condi tions. It is specific to plant conditions the procedure utilizes to make operational decisions.-Can the question be answered solely by knowing the purpose, overall sequenceof events, or overall mitigati ve strategy of a procedure? No, detailed and notoverall knowledge of steps and sequencing is required to answer the question.-Does the question re quire one or more of the following?* Assessing plant conditions (normal, abnormal, or emergency) and thenselecting a procedure or section of a procedure to mitigate, recover, orwith which to proceed Yes, the questi on requires the candidate to have a high level of understandi ng of the operational goals associated with adecision tree encompassed by 5 steps in AOP 18009-C which thendetermine the mitigating strategy that will be utilized.* Knowledge of when to impl ement attachments and appendices, including how to coordinate these it ems with procedure steps* Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific s ubprocedures or emergency contingency procedures* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant norma l, abnormal, and emergency proceduresFriday, February 21, 2014 3:43:48 PM 3
Level: SROTier # / Group # T2 / G2 K/A# 055G2.2.42 Importance Rating: 3.9 / 4.6 Technical
Reference:
TS 3.4.13, Rev Amendment No. 144, page 3.4.13-1 AOP 18009-C, Rev 29.2, page 6 & 7References provided: NoneLearning Objective: LO-TA-60020A Re spond to a Steam Generator Tube Leak per 18009-CLO-TA-16010 RCS Leakage Ca lculation (Inventory Balance) using 14905-1/2 LO-LP-39208-05 Given a set of Tech Specs and the bases, determine for a specific set of plant
conditions, equipment availability, and
operational mode:
- a. Whether any Tech Spec LCOs of section 3.4 are exceeded.
b.The required actions for all section 3.4 LCOs.LO-LP-39208-04 Descr ibe the bases for any given Tech Spec in section 3.4.
LO-LP-37311-02 Descr ibe the response of the following parameters to a Steam Generator Tube
Rupture: (include in the discussion the
response at power, during a reactor
startup, and after a reactor trip/safety injection)k. Steam jet ai r ejector and steam packing exhauster radiation monitor LO-LP-60309-10 Discuss how changes in the followingaffect radiation monitor response to a
steam generator tube leak/rupture:
- a. RCS activity
- b. Power level
- c. Process flow rate (i.e., SG blowdown)d. Rupture size Question origin: MODIFIED -
HL17 NRC Question # 37AA2.10Cognitive Level: M/F
10 CFR Part 55 Content: 41.5 / 43.5Comments: You have completed the test!Friday, February 21, 2014 3:43:48 PM 4
- 1. 037AA2.10 001/1/2/SGTL TECH SPEC/F 3.2/4.1/NEW/HL-17 NRC/SRO/EMT/GCW Unit 2 is experiencing a Steam Generator Tube Leak on SG
- 4. The crew is performing 18009-C, "Steam Ge nerator Tube Leak".
Current conditions:
- Reactor power is 100% and stable.
- 2RE-0724 N-16 Rad M onitor indicates 155 gpd.
- 2RE-0810 SJAE Exhaust Rad Monitor indicates 160 gpd. - RCS specific activity is 1.31 X 10
-3 micro Curies per gram DOSE EQUIVALENT I-131.
Based on these conditions, wh ich one of the following correctly completes the followingstatement?
Per Tech Spec 3.4.13, "RCS Operational Leakage", the primary to secondary leakageis ___(1)___ the limit and per the Tech Spec Bases 3.4.
17, "Steam Generator Tube Integrity," the limit ensures that under the stress of a LOCA or MSLB a single crack leaking this amount will not
___(2)___ .(1) within (2) exceed the limits for secondary coolant activity (1) within(2) propagate to a SGTR (1) exceeding (2) exceed the limits for secondary coolant activity (1) exceeding (2) propagate to a SGTR A.B.C.D.Monday, January 20, 2014 2:25:25 PM 1
RCS Operational LEAKAGE 3.4.13 Vogtle Units 1 and 2 3.4.13-1 Amendment No. 144 (Unit 1) Amendment No. 124 (Unit 2) 3.4 REACTOR COOLANT SYSTEM (RCS)
3.4.13 RCS Operational LEAKAGE
LCO 3.4.13 RCS operational LEAKAGE shall be limited to:
- a. No pressure boundary LEAKAGE;
- b. 1 gpm unidentified LEAKAGE;
- c. 10 gpm identified LEAKAGE; and
- d. 150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RCS operational LEAKAGE not within limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.
A.1 Reduce LEAKAGE to within limits.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> B. Required Action and associated Completion
Time of Condition A not met. OR Pressure boundary LEAKAGE exists.
OR Primary to secondary LEAKAGE not within limit. B.1 Be in MODE 3.
AND B.2 Be in MODE 5.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Approved By Vogtle Electric Generating Plant Procedure Version J Thomas 18009-C 29.2 Effective Date STEAM GENERA TOR TUBE LEAK Page Number 08/16/2012 6 of 34 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED Printed January 20, 2014 at 14:32 NOTE If available, both RE-0810 and RE-0724 should be used to determine the leakage rate of change in Step 7 RNO; however, if only o ne of the two radiat ion monitors is Functional, then the reading fr om the Functional monitor should be used to determine leakage rate of change. 6. Check Radiation monitors available:
RE-810 OR
RE-724 7. 6. Go to Step 8 6 7. Check leakage rate of change:
- 7. 7 a. Greater than or equal to 30 GPD/HR based on a 20 minute
trend: a. Perform the following:
7.a 1) After a 20 minute trend has elapsed, determine
the leakage rate of change. 7.a.1) IPC Points:
IF leakage rate of change is greater than or equal to 30 gpd/hr, THEN go to Step 8.
RE-0810: UR6810(GPD) UR6811(ROC)
RE-0724: UR6724(GPD) UR6725(ROC)
-OR- IF leakage rate of change is less than 30 gpd/hr, THEN go to Step 9.
S Approved By Vogtle Electric Generating Plant Procedure Version J Thomas 18009-C 29.2 Effective Date STEAM GENERA TOR TUBE LEAK Page Number 08/16/2012 7 of 34 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED Printed January 20, 2014 at 14:32
- 8. Check leakage rate - LESS THAN 75 GPD. 8. Perform the following:
8 a. Initiate 18013-C, RAPID POWER REDUCTION.
8.a b. Be less than 50% power within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
8.b c. Be in Mode 3 within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. 8.c d. Go to Step 12.
8.d 9. Check leakage rate - LESS THAN 150 GPD. 9. Perform the following:
9 a. Initiate 12004-C, POWER OPERATION (MODE 1).
9.a b. Be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
9.b c. Go to Step 12.
9.c 10. Check leakage rate - LESS THAN 75 GPD. 10. IF leakrate has remained greater than or equal to 75 gpd for one
- hour, THEN perform the following:
10 a. Initiate 12004-C, POWER OPERATION (MODE 1).
10.a b. Be in Mode 3 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 10.b c. Go to Step 12.
10.c S
- 1. 058AA2.03 001/LOIT AND LOCT/SRO/C/A 3.5/3.9/058AA2.03/LO-TA-60040A//HL15/
Procedure titles are as follows:
- 18034-1, "Loss of Class 1E 125 VDC Power" - 19000-C, "Reactor Trip or Safe ty Injection" Initial condition:
- Unit 1 is at 100% reactor power.
Current conditions:
- All Train 'A' MSIV re d and green handswitch lights extinguish. - RTB 'A' red and green lights extinguish.
- RCP #1 1E breaker re d and green handswitch lights extinguish.
- Channel I TSLB bistable lights illuminate.
Which one of the following completes the following statement?
The Shift Supervisor __(1)__
required to enter 18034-1, and the Shift Supervisor __(2)
__ required to enter 19000-C. __(1)__ __(2)__ is is is is NOT is NOT is is NOT is NOT A.B.C.D.K/A058 Loss of DC PowerAA2.03 Ability to determine and interpret the following as they apply to theLoss of DC Power: - DC loads lost; impact on ability to operate and monitor plantsystems.K/A MATCH ANALYSIS The question tests the candid ate's ability to relate mult iple indications and the immediate impact to plant operations. They must interp ret these various indicationsFriday, February 21, 2014 3:47:35 PM 1
and make a decision on which procedures would address the problem. EXPLANATION OF REQUIRED KNOWLEDGE The indications given are the symptoms of a loss of power to 125VDC bus 1AD1. All 'A' train 1E switchgear breakers will loose contro l power. As such, the breaker will remain in its current state without el ectrical protection and all han dswitch indication lights will be de-energized. The MSIV s and MFIVs will fail CLOSED as their solenoids are de-energized resulting in a Re actor Trip. All Channel I TSLBs lights will illuminate due to the loss of 1AY1A, which is normally feed from an inverter supplied by 1AD1.
Entry conditions for 18034-1 are met. Step 1 of 18034
-1 directs a reactor trip andINITIATION of 19000-C. Per Admin procedure 10020-C step 3.3, "Initiate" means the referenced procedure will be used as a supplement to, and it will be performed concurrently with the one in effect. Therefore, 18034-C and 19000-C are expected to be worked in conjunction due to the complications resulting from a loss of DC.
18034-1 will address all issues required by the loss of the supported 120V Vital AC bus.ANSWER / DISTRACTOR ANALYSISA. Correct. The first part is correct. Entry conditions for 18034-1 are met.
The second part is correct, Step 1 of 18034-1 directs a reactor trip and INITIATION of 19000-C.
B. Incorrect. Plausible. The first part is correct. Entry conditions for 18034-1 are met.
The second part is incorrect.
Per step 1 RNO of 18034-1, a reactor trip should have occu red and is required. However, step 1 is not an IOA. A candidate without specific knowledge of the procedure and w ho did not realize the MISIVs could conclude the reactor did not tr ip and entry into 19000-C is not required. Therefore, this distractor is plausible.
C. Incorrect. Plausible. The first part is incorrect. The loss of indication on the associated handswitches and TSLBs indicate a loss of 1AD1.
However, a candidate with insu fficient knowledge of power supplies could conclude that the handswitches are feed from 120V Vital AC and not the 125VDC supply. This is a common
misconception. As such, it would be reasonable for the candidate to conclude a loss of 1AY1A to have occurred and entry into 18034-1 is not required.
Therefore, this distractor is plausible.
The second part is correct. Step 1 of 18034-1 directs a reactor trip and INITIATION of 19000-C. Howeve r, a candidate that believes 18032-1 was entered in stead of 18034-1, would still find it reasonable that the reactor would trip based on othertrips. Therefore, this distractor is plausible.
D. Incorrect. Plausible. The first part is in correct. See the first part of choice C above.Friday, February 21, 2014 3:47:35 PM 2
The second part is incorrect.
Step 1 of 18034-1 directs areactor trip and INITIATION of 19000-C. Ho wever, a candidate that believes 18032-1 was entered instead of 18034-1 and has
knowledge of 18032-1, would no t expect the reactor to trip.SRO JUSTIFICATION (10CFR43(b))(5) Assessment of facility conditions and selection of appropriate proceduresduring normal, abnormal, and emergency situations.-Can the question be answered solely by knowing "systems knowledge", i.e.,how the system works, flowpath, logic, component location?
No, selection ofthe appropriate AOP/EOP combination is NOT associated with systemknowledge.-Can the question be answered solely by knowing immediate operator actions?
No, the direction to initiate 19000-C concurrent with 18034-C is the RNO of step 1 of 18034-C and is not an IOA.-Can the question be answered solely by knowing entry conditions for AOPs orplant parameters that require direct entry to major EOPs?
No, the entrycondition for each AOP/EOP is part of the question. However, the specificknowledge of 18034-C step 1 RNO is required. Normally, AOP use isdiscontinued upon entry in to 19000-C. Most AOPs th at direct tripping thereactor say "go to 19000-C", indicating that a transition to the EOP network ismade. There are only a select few AOPs that required concurrentimplementation of the AOP with the EOP. In most situation, the AOP is onlyperformed at SS discretion and is typically not needed. 19000-C is designed tobe successful with only one train of electrical power.-Can the question be answered solely by knowing the purpose, overall sequenceof events, or overall mitigative strategy of a procedure?
No, specific knowledgeof 18034-C step 1 RNO is required.-Does the question re quire one or more of the following?* Assessing plant conditions (normal, abnormal, or emergency) and thenselecting a procedure or section of a procedure to mitigate, recover, orwith which to proceed.
Yes, the candidate is required to select theappropriate procedure/combination of procedures to mitigate the plantconditions.* Knowledge of when to impl ement attachments and appendices, including how to coordinate these it ems with procedure steps* Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific s ubprocedures or emergency contingency procedures* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant norma l, abnormal, and emergency procedures Level: SROTier # / Group # T1 / G1 K/A# 058AA2.03 Importance Rating: 3.5 / 3.9Friday, February 21, 2014 3:47:36 PM 3
Technical
Reference:
AOP 18034-1 Rev 13.1, page 4 Admin Proc 10020-C Rev 9.0, page 8References provided: None
Learning Objective:
LO-LP-60329-01 Given that a loss of power has occurred to any of the following 125VDC vital buses
and given the appropriate plant
procedures, describe t he operator actions required and why thes e actions are taken.
- a. 1AD1 LO-LP-60329-04 Given condit ions and/or indications, determine the required AOP to enter (including subsection s, as applicable).LO-TA-60040A Respond to a Loss of Class 1E 125 VDC Power per 18034-1/2.Question origin: BANK - HL 15 NRC Question # 058AA2.03Cognitive Level: C/A
10 CFR Part 55 Content: 41.7 / 41.10 / 43.5Comments: Question appears to ma tch the KA. Not sure if the procedure question is at the SRO-only level. The secondquestion is systems knowledge and not at the SRO-only
level.Choices A-D have 4 differen t answers; the applicant does not need to know the answer to the second question to answer the question correctly.
The explanation states that an automatic reactor trip has occurred, however, t he justification for the correct answer is that the RNO step for "Verify Reac tor Trip," states to initiate 19000-C. Does this mean that, although it occurred automatically, Reactor Trip cannot be verified?
One possible fix could be droppi ng the second question (and anything in the stem that wa s solely used to answer the second question) and separat e the first question to ask:
"-the Shift Supervisor <is/is no t> required to enter 18034-1 and the Shift Supervisor required to enter 19000-C." (I still need to think about whether asking it this way is at the SRO-only level. I'm leaning to wards it IS at the SRO-only level.)
- JAT 12/19/2013 (Editorial)
New question incorporates t he above comments. SRO-onlyFriday, February 21, 2014 3:47:36 PM 4
appears to be met because the question requires knowledgeof procedure rules-of-usage.
- JAT 2/4/14You have completed the test!Friday, February 21, 2014 3:47:36 PM 5
Approved By Vogtle Electric Generating Plant Procedure Number Rev J. Thomas 18034-1 13.1 Date A pproved LOSS OF CLASS 1E 125V DC POWER Pa g e Numbe r 3/16/12 1 of 84 Printed January 21, 2014 at 12:44 ABNORMAL OPERATING PROCEDURE CONTINUOUS USE PURPOSE This procedure provides operator actions to be followed in the event that power is lost to one of the 125V DC Vital Busses (1AD1, 1BD1, 1CD1, or 1DD1).
Specific instructional steps will be found in the following sections:
A. LOSS OF 125V DC BUS 1AD1
B. LOSS OF 125V DC BUS 1BD1
C. LOSS OF 125V DC BUS 1CD1
D. LOSS OF 125V DC BUS 1DD1
SYMPTOMS SECTION A. LOSS OF 125V DC BUS 1AD1 125V DC Vital Bus 1AD1 voltage low. Loss of power to 1AY1A and 1AY2A 120V AC Vital Instrument Panels. Loss of indicating lights on 1AA02, 1AB04, 1AB05, and 1AB15 Switchgear Controls. Train A Main Steamline Isolation. Train A Main Feedwater Isolation.
SECTION B. LOSS OF 125V DC BUS 1BD1 125V DC Vital Bus 1BD1 voltage low. Loss of power to 1BY1B and 1BY2B 120V AC Vital Instrument Panels. Loss of indicating lights on 1BA03, 1BB06, 1BB07, and 1BB16 Switchgear Controls. Train B Main Steamline Isolation. Train B Main Feedwater Isolation.
Approved By Vogtle Electric Generating Plant Procedure Number Rev J. Thomas 18034-1 13.1 Date Approved LOSS OF CLASS 1E 125V DC POWER Page Number 3/16/12 4 of 84 A. LOSS OF 125V DC BUS 1AD1 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED Printed November 25, 2013 at 12:20 NOTES This procedure should be performed concu rrent with 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION. RCP 1 undervoltage and underfr equency trips will NOT actuate. See ATTACHMENT A for equipment responses, breaker and valve control loss, valve failures from loss of instrument air, and annunciator failures.
__A1. Verify reactor trip.
A1. Perform the following: A1 __a. Trip the reactor.
A1.a __b. Initiate 19000-C, E-0 REACTOR TRIP OR
SAFETY INJECTION.
A1.b __A2. Initiate the Continuous Actions Page.
A2. A2 __A3. Dispatch an operator to 1AA02 SWGR Room (CB-A48).
A3. A3 NOTE IF DG1A is NOT running, it can NOT be started.
__A4. Check DG1A - RUNNING.
__A4. Go to Step A7. A4 S Approved By C.S. WALDRUP Vogtle Electric Generating Plant Procedure Number Rev 10020-C 9 Date Approved 01/26/2011 EOP AND AOP RULES OF USAGE Page Number 8 of 27 Printed December 4, 2013 at 14:49 3.2 "GO TO" STEPS To maintain consistency in refer encing or branching to another procedure:
3.2.1 "Go to" is used when it is desired to branch to another procedure or to a preceding step in the procedure.
Example: IF the reactor trips, THEN go to 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION.
Branching implies the procedure in use shall be exited and a new procedure entered.
3.2.2 "return to" is used when it is desir ed to branch to a previous step in the procedure.
3.3 "BY INITIATING" STEPS When "by initiating" is used, t he referenced procedure will be used as a supplement to, and it will be performed c oncurrently with the one in effect.
3.4 IMMEDIATE
OPERATOR ACTIONS STEPS
3.4.1 These
are actions that, for EOPs ar e to be committed to memory for immediate performance upon initiation of the procedur
- e. These actions, which typically involve verification of automatic actions , are listed starting on top of the next page after the symptoms section with "IMMEDIATE OPERATOR ACTIONS" typed above Step 1.
3.4.2 Immediate
Operator Ac tion Steps shall be performed by memory by the operator.
The Unit Shift Supervisor will state the high level steps as written in the procedure. Upon restatement the operat or will repeat the step including all substeps to ensure completeness.
3.4.3 All EOP immediate actions must be comp leted prior to taking any early action or non-EOP action.
- 1. 062AG2.2.12 001/LOIT/SRO/C/A 3.7/2.7/062AG2.2.12/LO-LP-39204-04///
Initial condition:
- Unit 1 is at 100% reactor power.
Current conditions:
- The Shift Supervisor discovers the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Channel Check for Train 'A' NSCW basin level, 1LI
-1606, was missed.
- The last per formance of the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Channel Check was 0030 on 5-11-14.
- Time of disco very of the missed surve illance was 1700 on 5-12-14.- A risk evaluation will NOT be performed.
Which one of the following completes the following statement?
To prevent declaring the LC O NOT met, the surveillance is required to be performed satisfactorily no later than ________.
0030 on 5-12-14
0630 on 5-12-14 1700 on 5-13-14 0100 on 5-14-14 A.B.C.D.K/A062 Loss of NSCWG2.2.12 Knowledge of surveillance procedures
.K/A MATCH ANALYSIS The question tests the candidates knowled ge of generic TS survelliance SR 3.0.3 as specifically applied to a missed NSCW survelliance. If th e missed survelliance results in an inoperable declaration, a loss of o ne train of NSCW would occur since the survelliance affects the Ultim ate Heat Sink LCO 3.7.8.EXPLANATION OF REQUIRED KNOWLEDGE A Surveillance has been ident ified on an NSCW System as being missed and the candidate must use Tech Spec SR 3.0.3 to determine when the mi ssed surveillance isto be performed by. If it is discovered t hat a Surveillance was not performed within its specified Frequency, then compliance with the requirement to decla re the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whicheve r is greater. This delay period is permitted to allowMonday, February 24, 2014 9:43:58 AM 1
performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> s and the risk impact shall be managed. If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicabl e Condition(s) must be entered.ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible.
Incorrect but plausible because the candidate may determine the surveillance must be perform within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the time missed as opposed to time of discovery.
B. Incorrect. Plausible.
Incorrect but plausible because the candidate may determine the surveillance must be perfo rm from time missed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> plus the 125% grace period described in SR 3.0.2. The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from th e previous performance or as measured from the time a spec ified condition of the Frequency is met.C. Correct. The answer is correct the missed surveillance must be completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of poin t of discovery. (e.g. If it is discovered that a Surveillance was not performed within its specified Frequency, then comp liance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is greater.
This delay period is permitted to allow performance of the Surv eillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24
hours and the risk impact shall be managed. If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable
Condition(s) must be entered.)
D. Incorrect. Plausible. In correct but plausible because the candidate may determine the surveillance must be perform from point of discovery within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> plus the 125%
grace period described in SR 3.0.2.
The specified Frequency for each SR is met if the Surveillance
is performed within 1.25 times t he interval specified in the Frequency, as measured from th e previous performance or as measured from the time a spec ified condition of the Frequency is met.SRO JUSTIFICATION(2) Facility operating limitations in the technical specifications and their bases.-Can question be answered solely by knowing = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action?
No, thequestion is not related to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action time requirements.
-Can question be answered solely by knowing the LCO/TRM information listed "above-the-line?"
No, the question is not related to above-the-line information.Monday, February 24, 2014 9:43:58 AM 2
-Can question be answered solely by knowing the TS Safety Limits?
No, thequestion is not related to Safety Limits.-Does the question involve one or more of the following for TS,TRM, or ODCM?
- Application of Required Actions (Sec tion 3) and Surve illance Requirements (Section 4) in accordance with rules of applicati on requirements (Section 1)* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR4.0.1 thru 4.0.4)
Yes, the required knowledge is application of SR 3.0.3 in Tech Spec.* Knowledge of TS bases that is required to analyze TS required actions and terminology Level: SROTier # / Group # T1 / G1 K/A# 062G2.2.12 Importance Rating: 3.7 / 4.1Technical
Reference:
OS P 14000-1, Rev 88.1, page 15 TS SR 3.0.3, Amendm ent No. 125, page 3.0-4 Surv Frequency Control Program, Rev 3, page 15References provided: None
Learning Objective:
LO-LP-39204-04 State the allo wable time intervals for extension of surveillances. State the
result of failure to perform surveillances within this period.
LO-LP-39204-06 In regard to surveillances, determine when time delay may be applied and the maximum time allowed to perform the surveillance.Question origin: BANK Cognitive Level: C/A
10 CFR Part 55 Content: 41.10 / 43.2Comments: You have completed the test!Monday, February 24, 2014 9:43:58 AM 3
SR Applicability
3.0 Vogtle
Units 1 and 2 3.0-4 Amendment No. 125 (Unit 1)
Amendment No. 103 (Unit 2) 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3. Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.
SR 3.0.2 The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.
For Frequencies specified as "once," the above interval extension does not apply.
If a Completion Time requires periodic performance on a "once per . . ."
basis, the above Frequency extension applies to each performance after the initial performance.
Exceptions to this Specification are stated in the individual Specifications.
SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the risk impact shall be managed.
If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.
(continued)
VEGP - Surveillance Frequency Control Program Page 15 of 19 Revision 3 SurveillanceRequirement Frequency Notes LDCR No. SR 3.7.7.2 18 months on a STAGGERED TEST BASIS 2012-015 SR 3.7.8.1 31 days N/A SR 3.7.8.2 18 months For the following components only:
1HV1668A 2HV1668A 1HV1668B 2HV1668B 1HV1669A 2HV1669A 1HV1669B 2HV1669B N/A 18 months on a STAGGERED TEST BASIS For the following components only:
1HV1806 2HV1806 1HV1808 2HV1808 1HV1822 2HV1822 1HV1830 2HV1830 1HV2134 2HV2134 1HV2138 2HV2138 1HV1807 2HV1807 1HV1809 2HV1809 1HV1823 2HV1823 1HV1831 2HV1831 1HV2135 2HV2135 1HV2138 2HV2138 2012-015 SR 3.7.8.3 18 months on a STAGGERED TEST BASIS 2012-015 SR 3.7.9.1 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> N/A SR 3.7.9.2 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> N/A SR 3.7.9.3 31 days N/A SR 3.7.9.5 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> N/A Approved By J.B. Stanley Vogtle Electric Generating Plant Procedure Version 14000-1 88.1 Effective Date 06/21/2013 OPERATIONS SHIFT AND DAILY SURVEILLANCE LOGS Page Number 15 of 36 Printed January 3, 2014 at 14:02 Sheet 9 of 10 DATA SHEET 1 MODE 1 & 2 MODE _______________
DATE _______________
LCO TECH SPEC INDICATION LIMIT(S)
METHOD OF VERIFICATION SURV REQ PARAMETER INSTRUMENT DAY NIGHT TOLERANCE LCO/PROC CREFS ACTUATION OPERABLE SR 3.3.7.1
FCN 3 CR INTAKE RADIATION 1RE-12116 CHANNEL CHECK
3.3.7 CHANNEL
CHECK MONITORS (INIT) 1RE-12117 REQUIRED 2 FHB ACTUATION OPERABLE TRS 13.3.6.1 FHB EFFL RADIOGAS ARE-2532A
- 13.3.6 CHANNEL CHECK FHB ISO (INIT) ARE-2532B REQUIRED 1 FHB ACTUATION OPERABLE TRS 13.3.6.1 FHB EFFL RADIOGAS ARE-2533A
- 13.3.6 CHANNEL CHECK FHB ISO (INIT) ARE-2533B REQUIRED 1 *INDICATING NORMALLY. ALL STATUS AND ALARM LIGHTS EXTINGUISHED. DG1A FUEL OIL INVENTORY VERIFY FUEL OIL STORAGE TANK LEVEL SR 3.8.3.1 DG 1A LEVEL (%) 1-LI-9024 82% 3.8.3 DG1B FUEL OIL INVENTORY VERIFY FUEL OIL STORAGE TANK LEVEL SR 3.8.3.1 DG 1B LEVEL (%) 1-LI-9025 82% 3.8.3 TWO INDEPENDENT CONTROL ROOM EMERGENCY FILTRATION SYSTEMS SHALL BE OPERABLE VERIFY CONTROL ROOM TEMP SR 3.7.10.1 SR 3.7.11.1
NOTE: TEMPERATURE INDICATION IS OBTAINED FROM HAND-HELD TEST EQUIPMENT. RECORD INSTRUMENT INFORMATION BELOW.
INSTRUMENT ID NO.
N/A CAL DUE DATE CONTROL ROOM TEMPERATURE (F) M&TE <85 F 3.7.10 3.7.11 THE RWST SHALL BE OPERABLE VERIFY TEMPERATURE SR 3.5.4.1 TRS 13.1.7.1 RWST TEMPERATURE (F) 1TIS-10980
>51 F * <109 F
- 3.5.4 13.1.7 *WITH INDICATED RWST TEMPERATURE OUTSIDE THE LIMITS, THEN VERIFY RWST TEMPERATURE IS WITHIN TECHNICAL SPECIFICATION LIMITS BY PLACING THE RWST ON RECIRC USING SLUDGE MIXING PUMP WITH HEATER OFF AND OBSERVING 1-TI-10982 TO BE WITHIN 44F AND 116F. THE ULTIMATE HEAT SINK SHALL BE OPERABLE COMPUTER POINT T2601* <90 F 3.7.9 VERIFY WATER -OR- TEMPERATURE AND LEVEL SR 3.7.9.2 TEMPERATURE (F) 1TJI-1692 POINT 2* COMPUTER POINT T2602* -OR- 1TJI-1692 POINT 17* *IF COMPUTER POINT AND RECORDER POINT ARE NOT AVAILABLE, TEMPERATURE READING MUST BE OBTAINED LOCALLY USING HAND-HELD TEST EQUIPMENT. RECORD INSTRUMENT INFORMATION BELOW.
INSTRUMENT ID NO.
N/A CAL DUE DATE SR 3.7.9.1 LEVEL 1LI-1606 >73% (%) 1LI-1607 CONTAINMENT AIR TEMPERATURE SHALL NOT SR 3.6.5.1 COMPUTER POINT T2501 EXCEED 120F VERIFY AVERAGE AIR TEMPERATURE (F) COMPUTER POINT T2502 NA TEMPERATURE COMPUTER POINT T2503 COMPUTER POINT UT2501 (AVG)
<120 F 3.6.5 *IF COMPUTER POINT IS NOT AVAILABLE VERIFY CNMT HI TEMP ALARM ALB-01 (E06) IS NOT IN ALARM. ALB-01 (E06) NOT IN ALARM
- IF COMPUTER POINT AND ALB-01 (E06) ARE NOT AVAILABLE, TEMPERATURE READING MUST BE OBTAINED LOCALLY USING HAND-HELD TEST EQUIPMENT FOR 1TE-2612 FOR POINT T2502 AND 1TE-2613. FOR POINT T2503 RECORD INSTRUMENT INFORMATION BELOW. USE MCB INDICATOR 1TI-2563 FOR POINT T2501 AND AVERAGE THE THREE.
INSTRUMENT ID NO.
<120 F CAL DUE DATE COMPLETED BY: DAY: TIME: NIGHT: TIME: SS REVIEW: DAY: TIME: NIGHT: TIME:
- 1. 076A2.02 001/LOIT AND LOCT/SRO/C/A 2.7/3.1/076A2.02/LO-TA-60003/63013///
Initial conditions:
- Unit 1 is at 100% reactor power.
- SI Pump 'A' is tagged out.
Current conditions:
- Train 'B' NSCW supply heade r pressure is 75 ps ig and lowering.
- Train 'B' NSCW s upply header flow is 25,000 gpm. - Train 'B' NSCW re turn header flow is 10,000 gpm.
- 18021-C, "Loss of Nuclear Service Cooling Water System," is entered.
Which one of the following completes the following statement?
Per 18021-C, the crew is required to __(1)__ the stan dby Train 'B' NSCW pump, and after completing the actions of 18021-C, the Shift S upervisor will determine per 10008-C, "Recording Limiting Conditions for Operation," that a LOSF __(2)__ exist.__(1)__ __(2)__ start does start does NOT place in PTL does place in PTL does NOT A.B.C.D.K/A 076 Service Water A2.02 Ability to (a) predict the impacts of the following malfunctions oroperations on the SWS; and (b) based on those predictions, useprocedures to correct, control, or mitigate the consequences ofthose malfunctions or operations: - Service water header pressureK/A MATCH ANALYSIS The question addresses a problem identif ied on an NSCW System to include lowMonday, February 24, 2014 2:02:36 PM 1
header pressure and the candid ate must determine the corre ct procedure action based on the indications provided.
In addition, the candidates must determine the impact on plant operations using the in formation provided in the stem as related to LOSF evaluation. This concept brings t he question to the SRO knowledge level.EXPLANATION OF REQUIRED KNOWLEDGE Per 18021-C symptoms, a drop in NSCW heade r pressure accompanied by a large difference between supply and return header flow s indicates a large (c atastrophic) leak.
Per steps 1 and 2, all pumps in the affe cted train will be placed in PTL.Since NSCW 'B' is a support system for all 'B' train ECCS pumps, a LOSF function exists with SIPs. With SIP 'A' tagged out in the stem and a subsequent loss of SIP 'B'due to loss of a support system, no medium head injection is available. Since TS 3.5.2 does not have a condition for two trains of ECCS inoperable, TS 3.
0.3 must be entered. (Reference 10008-C for LOSF evaluation guidance.)ANSWER / DISTRACTOR ANALYSISA. Incorrect. Plausible. Part 1 is incorr ect however 'plausible' since the candidate may believe the stem indications are due a pump problem such as a broken coupling, as opposed to a leak since the symptoms would be the same wit h exception of the supply and return flow deviation. Per 18021-C step 6, the correct action for this condition would be to start the standby pump.
Part 2 is correct. The candidat e should determine that with both Train 'A' SI Pump and Train 'B
' NSCW System inoperable aLOSF exists per 10008-C 'Recor ding Limiting Conditions for Operation'. However, if the candidate misses the catastrophic leak in Part 1, it would still be plausible for them to incorrectly determine that the st andby NSCW pump is inoperable due to the failure to start on low h eader pressure and determine a LOSF exists for the wrong reason.B. Incorrect. Plausible. Part 1 is incorrect. See Part 1 of choice A above. Part 2 is incorrect. The cand idate should dete rmine that with both Train 'A' SI Pump and Train 'B' NSCW System inoperable a LOSF exists per 10008-C 'Recor ding Limiting C onditions for Operation'. However, if the candidate misses the catastrophic leak in Part 1 and recognizes that the failure of the start on low header pressure in not a requir ed function, then determining that a LOSF does not exist would be correct for these erroneous conditions..C. Correct. Part 1 is correct. The ca ndidate should determi ne that from the stem information prov ided that a large leak has occurred in the NSCW piping and the correct actions per 18021-C 'Loss of NSCW', is to place th e affected pumps in PTL.Monday, February 24, 2014 2:02:36 PM 2
Part 2 is correct. The candidat e should determine that with both Train 'A' SI Pump and Train 'B
' NSCW System inoperable aLOSF exists per 10008-C 'Recor ding Limiting Conditions for Operation'.
D. Incorrect. Plausible. Part 1 is correct. See Part 1 of choice C above.
Part 2 is incorrect however 'pl ausible' since the candidate maynot make the operability connection between the two systems
or assume that NSCW could be run in single pump operations to supply cooling and therefore be lieve a LOSF condition is not present.SRO JUSTIFICATION(2) Facility operating limitations in the technical specifications and their bases.-Can question be answered solely by knowing = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action? No, thequestion only addresses TS act ion of 72 hrs in association with implementation of TS 3.0.3.
-Can question be answered solely by knowing the LCO/TRM information listed "above-the-line?" No, all necessary knowledge is below the line.-Can question be answered solely by knowing the TS Safety Limits?
No, the question does not address Safety Limits.-Does the question involve one or more of the following for TS,TRM, or ODCM?
- Application of Required Actions (Sec tion 3) and Surve illance Requirements (Section 4) in accordance with rules of applicati on requirements (Section 1)* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR4.0.1 thru 4.0.4) Yes, the candidate is required to determine theapplicability of LCO 3.0.3 as applied to a LOSF.* Knowledge of TS bases that is required to analyze TS required actions and terminology(5) Assessment of facility conditions and selection of appropriate proceduresduring normal, abnormal, and emergency situations.-Can the question be answered solely by knowing "systems knowledge", i.e.,how the system works, flowpath, logic, component location? No, systemknowledge will not address the operability/safety function determination.-Can the question be answered solely by knowing immediate operator actions? No IOA's are address ed by the question.-Can the question be answered solely by knowing entry conditions for AOPs orplant parameters that require direct entry to major EOPs? No, knowledge of anadministrative process is required associated with Tech Specs.-Can the question be answered solely by knowing the purpose, overall sequenceof events, or overall mitigati ve strategy of a procedure? No, sequen ce or overallstategy of 18021-C will not answer either part of the question.-Does the question re quire one or more of the following?* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or se ction of a procedure to mi tigate, recover, or withMonday, February 24, 2014 2:02:36 PM 3
which to proceed* Knowledge of when to impl ement attachments and appendices, including how to coordinate these it ems with procedure steps* Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific s ubprocedures or emergency contingency procedures* Knowledge of administrative procedures that specify hierarchy,implementation, and/or coordination of plant normal, abnormal, and emergency procedures Yes, specific knowle dge of Admin procedure10008-C is required to perform a LOSF evaluation to determine whichLCO(s) are not met based on the inoperability of support systems and theimpact to pre-existing inoperabilities.
Level: SROTier # / Group # T2 / G1 K/A# 076A2.02 Importance Rating: 2.7 / 3.1Technical
Reference:
ADMIN 10008-C, Rev 30.0, pages 1-6, 15-19, & 28 AOP 18021-C, Rev 19.0, pages 1-3
TS 3.5.2, Amendment No. 136, page 3.5.2-1References provided: None Learning Objective: LO-LP-63508-04 Define the follo wing terms per 10008-C.
- a. LCO
- d. Loss of Safety Function
- e. Supported System
- f. Support System LO-PP-06101-04 Describe the in dications of the following:
- c. Supply header leak
- d. Return header leak
- e. NSCW pump trip
- f. NSCW piping leak in a pump roomLO-TA-63013 Implement Tec hnical Specification LCO using 10008-C (SRO Only)LO-TA-60003 Respond to a Lo ss of NSCW per 18021-CQuestion origin: MODIFIED - HL15 Question # 062AA2.02Cognitive Level: C/A 10 CFR Part 55 Content: 41.4 / 41.10 / 43.2 / 43.5Comments: You have completed the test!Monday, February 24, 2014 2:02:36 PM 4
- 1. 062AA2.02 001/1/1/LOSS OF NSCW/C/A - 3.6/NEW/HL15/SRO/DS/TNT Given the following conditions at 38% power:- ACCW pump 2 is in service- CCW pumps 2 &
4 are in service- NSCW Pump 5 is danger taggedTrain A NSCW indications:
Train B NSCW indications:- Supply header pressure 45 psig
- Supply header pressure 58 psig- Supply header flow 8,000 gpm
- Supply header flow 25,000 gpm- Return header flow 8,0 00 gpm
- Return heade r flow 10,000 gpm Which of the following choi ces contains the correct pr ocedural entry and actions?
Enter AOP 18021-C, Loss of NSCW, due to loss of both NSCW Trains.
Place all NSCW pumps in PT L, trip the reactor and init iate EOP 19000-C. Trip the RCPs and isolate CVCS letdown.
Enter AOP 18021-C, Loss of NSCW, due to leakage on Train A NSCW.Place all Train A NSCW pumps in PTL, trip the reactor and initiate EOP 19000-C.
Trip the RCPs and isolate CV CS letdown if cooling not restored in 10 minutes.
Enter AOP 18021-C, Loss of NSCW, due to leakage on Train B NSCW.Place all Train B NSCW pumps in PTL.
Shift to Train A CCW pumps. Start ACCW pump #1 and remain in 18021-C.
Enter AOP 18021-C, Loss of NSCW, due to loss of both NSCW Trains.
Place NSCW Train B in single pump operation and a ll Train A NSCW pumps in PTL. Trip RCPs if seal temperatur es exceed 230 F and remain in 18021-C.
A.B.C.D.Tuesday, January 21, 2014 1:35:43 PM 1
ECCS-Operating
3.5.2 Vogtle
Units 1 and 2 3.5.2-1 Amendment No. 136 (Unit 1) Amendment No. 115 (Unit 2) 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)
3.5.2 ECCS - Operating
LCO 3.5.2 Two ECCS trains shall be OPERABLE.
APPLICABILITY: MODES 1, 2, and 3.
NOT E---------------------------------------------- In MODE 3, either residual heat removal pump to cold legs injection flow path may be isolated by closing the isolation valve to perform pressure isolation valve testing per SR 3.4.14.1.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more trains inoperable.
AND At least 100% of the ECCS flow equivalent to
a single OPERABLE
ECCS train available.
A.1 Restore train(s) to OPERABLE status.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B. Required Action and associated Completion Time not met.
B.1 Be in MODE 3.
AND B.2 Be in MODE 4.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
Approved By Vogtle Electric Generating Plant Procedure Version J. B. Stanley 18021-C 19 Effective Date LOSS OF NUCLEAR SERVICE COOLING WATER SYSTEM Pa g e Numbe r 11/09/2012 1 of 15 Printed October 3, 2013 at 12:28 ABNORMAL OPERATING PROCEDURE CONTINUOUS USE PURPOSE This procedure addresses the lo ss or degraded operation of one or more trains of Nuclear Service Cooling Water.
SYMPTOMS Trip of operating NSCW pumps and failure of standby pump to start. Dropping NSCW Supply Header pressure. Large difference between Supply Header flow and Return Header flow, indicating a large leak. NSCW Tower Basin temperature rising above 90°F. High temperature or low flow alarms on any components or systems cooled by NSCW.
MAJOR ACTIONS Determine condition causing loss or degraded operation of NSCW. Transfer loads to unaffected train. Correct or repair condition causin g loss or degraded operation of NSCW.
__
Approved By Vogtle Electric Generating Plant Procedure Version J. B. Stanley 18021-C 19 Effective Date LOSS OF NUCLEAR SERVICE COOLING WATER SYSTEM Page Number 11/09/2012 2 of 15 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED Printed October 3, 2013 at 12:28
- 1. Check if catastrophic leakage from NSCW system - EXISTS.
- 1. Go to Step 6.
1 2. Place affected train NSCW pump handswitches in PULL-TO-LOCK.
- 2. 2 3. Depress both Emergency Stop pushbuttons for the affected DG.
- 3. 3 4. Verify proper operation of UNAFFECTED NSCW train:
- 4. IF neither NSCW train can be placed in normal, two pump operation, THEN perform the following:
4 Two pumps running.
- a. Trip the reactor.
4.a Supply header pressure greater than 70 psig:
- b. Initiate 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION.
4.b Train A: PI-1636 Train B: PI-1637
- c. Trip all reactor coolant pumps.
4.c Supply header temperature computer indication less than
90°F: d. Isolate letdown.
4.d Train A: T2601 Train B: T2602
- e. Place one train of NSCW in single pump operation by initiating 13150, NUCLEAR SERVICE COOLING WATER SYSTEM. 4.e Supply header flow approximately 17,000 gpm:
Train A: FI-1640B Train B: FI-1641B
established using 13006, CHEMICAL AND VOLUME CONTROL SYSTEM.
4.f Step 4 continued on next page
Approved By Vogtle Electric Generating Plant Procedure Version J. B. Stanley 18021-C 19 Effective Date LOSS OF NUCLEAR SERVICE COOLING WATER SYSTEM Pa g e Numbe r 11/09/2012 1 of 15 Printed October 3, 2013 at 13:06 ABNORMAL OPERATING PROCEDURE CONTINUOUS USE PURPOSE This procedure addresses the lo ss or degraded operation of one or more trains of Nuclear Service Cooling Water.
SYMPTOMS Trip of operating NSCW pumps and failure of standby pump to start. Dropping NSCW Supply Header pressure. Large difference between Supply Header flow and Return Header flow, indicating a large leak. NSCW Tower Basin temperature rising above 90°F. High temperature or low flow alarms on any components or systems cooled by NSCW.
MAJOR ACTIONS Determine condition causing loss or degraded operation of NSCW. Transfer loads to unaffected train. Correct or repair condition causin g loss or degraded operation of NSCW.
__
Approved By Vogtle Electric Generating Plant Procedure Version J. B. Stanley 18021-C 19 Effective Date LOSS OF NUCLEAR SERVICE COOLING WATER SYSTEM Page Number 11/09/2012 2 of 15 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED Printed October 3, 2013 at 13:06
- 1. Check if catastrophic leakage from NSCW system - EXISTS.
- 1. Go to Step 6.
1 2. Place affected train NSCW pump handswitches in PULL-TO-LOCK.
- 2. 2 3. Depress both Emergency Stop pushbuttons for the affected DG.
- 3. 3 4. Verify proper operation of UNAFFECTED NSCW train:
- 4. IF neither NSCW train can be placed in normal, two pump operation, THEN perform the following:
4 Two pumps running.
- a. Trip the reactor.
4.a Supply header pressure greater than 70 psig:
- b. Initiate 19000-C, E-0 REACTOR TRIP OR SAFETY INJECTION.
4.b Train A: PI-1636 Train B: PI-1637
- c. Trip all reactor coolant pumps.
4.c Supply header temperature computer indication less than
90°F: d. Isolate letdown.
4.d Train A: T2601 Train B: T2602
- e. Place one train of NSCW in single pump operation by initiating 13150, NUCLEAR SERVICE COOLING WATER SYSTEM. 4.e Supply header flow approximately 17,000 gpm:
Train A: FI-1640B Train B: FI-1641B
established using 13006, CHEMICAL AND VOLUME CONTROL SYSTEM.
4.f Step 4 continued on next page
Approved By Vogtle Electric Generating Plant Procedure Version J. B. Stanley 18021-C 19 Effective Date LOSS OF NUCLEAR SERVICE COOLING WATER SYSTEM Page Number 11/09/2012 3 of 15 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED Printed October 3, 2013 at 13:06 g. Check RCP No. 1 seal temperatures less than 220°F.
4.g h. IF RCP No. 1 seal temperatures greater than 220°F, THEN do NOT attempt to restart RCPs prior to a status
evaluation.
4.h 5. Go to Step 13.
- 5. 5 6. Verify two or more NSCW pumps on the affected train are operating
properly by checking the following
parameters exist:
- 6. Perform the following:
6 Supply header pressure greater than 70 psig.
- a. Place affected train NSCW pump handswitches in
PULL-TO-LOCK.
6.a Train A: PI-1636 Train B: PI-1637
- b. Depress both Emergency Stop pushbuttons for the affected
DG. 6.b Supply header flow approximately 17,000 gpm.
- c. Investigate cause for trip of running pump(s).
6.c Train A: FI-1640B Train B: FI-1641B Step 6 continued on next page
Approved By J.B. Stanley Vogtle Electric Generating Plant Procedure Version 10008-C 30 Effective Date 02/08/2013 RECORDING LIMITING CO NDITIONS FOR OPERATION Page Number 1 of 32 Printed October 3, 2013 at 12:42
RECORDING LIMITING CONDITIONS FOR OPERATION
PROCEDURE LEVEL OF USE CLASSIFICATION PER NMP-AP-003 CATEGORY SECTIONS Continuous: NONE
Reference:
NONE Information: ALL
Approved By J.B. Stanley Vogtle Electric Generating Plant Procedure Version 10008-C 30 Effective Date 02/08/2013 RECORDING LIMITING CO NDITIONS FOR OPERATION Page Number 2 of 32 Printed October 3, 2013 at 12:42 TABLE OF CONTENTS PAGE 1.0PURPOSE 32.0PRECAUTIONS AND LIMITATIONS 33.0DEFINITIONS 43.1EXTENT OF CONDITION REVIEW 43.2LIMITING CONDITION FOR OPERATION (LCO) 43.3TECHNICAL REQUIREMENT (TR) 43.4INFORMATION ONLY LIMITING CO NDITION FOR OPERATION/TECHNICAL REQUIREMENT (Info LCO/TR) 53.5LOSS OF SAFETY FUNCTION (LOSF) 53.6SUPPORT SYSTEM 53.7SUPPORTED SYSTEM 64.0PROCEDURE 74.1LCO/TR STATUS SHEET PREPARATION 74.1.1Initiation Of LCO/TR Status Sheet For An LCO/TR 84.1.2Initiation Of LCO/TR Status Sheet For An Info LCO/TR 114.1.3Restoration Of LCO/TR s And Info LCO/TR s 124.2CONVERSION OF LCO/TR(S) TO INFO LCO/TR(S) 144.3CONVERSION OF INFO LCO/TR(s) TO LCO/TR(s) 144.4LCO/TR STATUS BINDER 144.4.1Part I. LCO/TR Status Log 144.4.2Part II. Active LCO Status Sheets 144.4.3Part III. Completed LCO/TR Status Sheets 154.5LOSS OF SAFETY FUNCTION (LOSF) EVALUATION 155.0RECORDS 2
06.0REFERENCES
20 Approved By J.B. Stanley Vogtle Electric Generating Plant Procedure Version 10008-C 30 Effective Date 02/08/2013 RECORDING LIMITING CO NDITIONS FOR OPERATION Page Number 3 of 32 Printed October 3, 2013 at 12:42
1.0 PURPOSE
1.1 This procedure prescribes the method to record the failure to meet the Limiting Conditions for Operation (LCO), or Technical Requirement, the associated ACTION requirements, any change in st atus effecting the ACTION, and the return to compliance with LCO/TR.
1.2 This procedure also includes inst ructions for implementing Technical Specification 5.5.15, the Safety Functi on Determination Program. As required by LCO 3.0.6, this program ensures t hat proper actions are taken such that multiple inoperable Structures, Systems, or Components (SSC) do not result in an undetected LOSS OF SAFETY FUNCTION.
1.3 This procedure also ensures that the allowed out of service time of SUPPORTED SYSTEMS is not inappropriately extended as a result of multiple inoperable SUPPORT SYSTEMS.
2.0 PRECAUTIONS
AND LIMITATIONS
2.1 Technical
Specification LCO 3.0.2 stat es that the required Actions of an LCO MUST be performed when the requirements of the LCO are NOT met. LCO
3.0.6 provides
an exception to LCO 3.0.2 for SUPPORTED SYSTEMS by NOT requiring the Required Actions for the SUPPORTED SYSTEMS to be performed WHEN the failure to meet an LCO is SOLELY due to the inoperability of a SUPPORT SYSTEM. In th is situation, LCO 3.0.6 requires ONLY the Required Actions of the SUPPORT SYSTEM to be performed. Since "cascading" is NOT required in this ca se, a possibility exists that unrelated concurrent failures of more than one SUPPORT SYSTEM could result in the complete loss of both trains of a SUPPORTED SYSTEM. THEREFORE, upon a failure to meet two or more LCOs during the same time period, an evaluation SHALL be conducted to determine if a LOSS OF SAFETY FUNCTION (LOSF) exists. This LOSF Evaluation satisfies t he criteria of Technical Specification Administrative Control 5.5.15, Sa fety Function Determination Program.
2.2 If the failure of a SSC not addressed by Technical Specification results in the inoperability of a required SUPPORT and/or SUPPORTED SYSTEM, then the LCO(s) for the required SUPPORT and/or SUPPORTED SYSTEM would be entered. Example: If 1NB10 is de-energiz ed, the operability of D/G '1B' or the Pressurizer Heaters may be impacted.
2.3 A single component inoperability can resu lt in multiple inoperabilities within a single train and affect multiple Technica l Specification LCOs. LCO 3.0.6 limits the amount of "cascading" of actions that is required when an inoperable SSC renders a SUPPORT SYSTEM inoperable.
Approved By J.B. Stanley Vogtle Electric Generating Plant Procedure Version 10008-C 30 Effective Date 02/08/2013 RECORDING LIMITING CO NDITIONS FOR OPERATION Page Number 4 of 32 Printed October 3, 2013 at 12:42 2.4 A single component inoperability CA N also impact operability on redundant trains. Example: IF 1HV-8716A is cl osed, both trains of ECCS may be impacted.
2.5 A Loss of Safety Function evaluati on must be performed for each inoperability of a SCC impacting a required SU PPORT or SUPPORTED SYSTEM(s).
2.6 The LOSF Evaluation MUST be reinitiated whenever an additional required structure, system, or component (SSC) is declared inoperable. This includes LCOs with Required Actions that s pecify declaring additional components inoperable.
2.7 Alternating
between LCO Conditions, in order to allow indefinite continued operation while not meeting the LCO, is not allowed.
3.0 DEFINITIONS
3.1 EXTENT
OF CONDITION REVIEW A review to determine the scope of SSCs (in other trains, units, or subcomponents) that are affected by a condition adverse to quality.
3.2 LIMITING
CONDITION FOR OPERATION (LCO)
A condition specified in the plant Techni cal Specifications (TS) or Technical Requirements Manual (TRM) which limits unit operations. An LCO may be
entered by an equipment malfunction or a change in a unit parameter. If an LCO is not met, the associated ACTION requirements shall be met.
3.3 TECHNICAL
REQUIREMENT (TR)
A condition specified in the plant Technical Requirements Manual (TRM) which limits unit operations. A TR may be enter ed by an equipment malfunction OR a change in a unit parameter. If a TR is not met, the associated ACTION
requirements SHALL be met. TR and Technical Requirement Surveillances (TRS) associated with each TR are implemented in the same way as Technical
Specifications. However, TRs and TRSs are treated as plant procedures and are not part of the Technical Specif ication. Therefore exceptions apply (Reference TRM Section 11.5).
Approved By J.B. Stanley Vogtle Electric Generating Plant Procedure Version 10008-C 30 Effective Date 02/08/2013 RECORDING LIMITING CO NDITIONS FOR OPERATION Page Number 5 of 32 Printed October 3, 2013 at 12:42
3.4 INFORMATION
ONLY LIMITING CONDITION FOR OPERATION/TECHNICAL REQUIREMENT (Info LCO/TR)
A method of tracking an equipment malf unction or change in plant parameter which would restrict unit operation in another mode OR prevent a mode change in which it would become applicable, or may become an LCO/TR for the present mode should other Technical Specific ation related equipment or redundant safety related equipment become inoperable.
Information Only LCOs should NOT be prepared for conditions that are not applicable in the present operating mode unless used for tracking for entry into
a mode in which a transition is to be directly made.
In addition, as an administrative tool to help track compliance with the ODCM, Information LCOs WILL be used when the requirements of the Offsite Dose Calculation Manual (ODCM) Sections 2.5 and 3.5 are not met.
Log entry (Electronic Log) of Information Only LCOs should NOT be made.
3.5 LOSS OF SAFETY FUNCTION (LOSF)
A LOSF exists WHEN; assuming no concu rrent single failure, a safety function assumed in the accident analysis cannot be performed.
3.6 SUPPORT
SYSTEM 3.6.1 A SSC which is needed by another Tec hnical Specification LCO required SSC to perform a safety function.
Example: The Component Cooling Wa ter System (SUPPORT SYSTEM) is required by the Residual Heat Removal System (SUPPORTED SYSTEM) to
fulfill its safety function. A SUPPORT SYSTEM may also be a SUPPORTED
SYSTEM. Example: The Component Coo ling Water System requires the Nuclear Service Cooling Water System to fulfill its safety function.
Approved By J.B. Stanley Vogtle Electric Generating Plant Procedure Version 10008-C 30 Effective Date 02/08/2013 RECORDING LIMITING CO NDITIONS FOR OPERATION Page Number 6 of 32 Printed October 3, 2013 at 12:42 3.6.2 For the purpose of implementing LCO 3.
0.6, a SSC which monitors or maintains a process parameter or operating limit is not a SUPPORT SYSTEM; however, specific functions of Technical Specificat ion instrumentation required to fulfill a credited safety function, may be considered a SUPPORT SYSTEM.
Examples:
The Digital Rod Position Indicators (DRPI) are used to monitor control rod insertion limits, however; inoperab ility of DRPI does not result in the control rods not being within insertion limits. Control r od insertion limits are monitored separately and actions taken as appropriate when
insertion limits are not met or Su rveillance Requirements not performed when required.
Likewise, parameter limit s that could affect ot her parameter limits if exceeded are also NOT considered SUPPORT SYSTEMS for the
purposes of implementing LCO 3.0.6.
Example: Exceeding control rod insertion limits could affect hot channel factors Auto Actuation Logic and Actuati on Relays, although identified as Instrumentation, MAY be c onsidered a SUPPORT SYSTEM.
3.7 SUPPORTED
SYSTEM A SSC, required by the Technical Spec ifications, which requires a SUPPORT SYSTEM to ensure its safety function c an be performed. For the purposes of implementing LCO 3.0.6, pr ocess parameters, operati ng limits, or individual instrument channels are NOT SUPPORT ED SYSTEMS; however, specific functions of Technical Specification inst rumentation required to fulfill a credited safety function, MAY be considered a SUPPORTED SYSTEM.
Approved By J.B. Stanley Vogtle Electric Generating Plant Procedure Version 10008-C 30 Effective Date 02/08/2013 RECORDING LIMITING CO NDITIONS FOR OPERATION Page Number 15 of 32 Printed October 3, 2013 at 12:42 4.4.3 Part III. Completed LCO/TR Status Sheets Part III contains copies of LCO/TR Status Sheets for LCO/TRs that have been restored. Sheets are filed in order of their LCO/TR number. Copies of completed sheets SHOULD be retained in the binder for at least 30 days after they have been closed out.
4.5 LOSS OF SAFETY FUNCTION (LOSF) EVALUATION
4.5.1 Review
Precautions and Limitations PRIOR to performing next step.
4.5.2 Identify
the applicable Technical Specificat ion conditions and required actions for the inoperable SSCs PRIOR to entering the LCO, IF possible.
NOTE A flow chart of the LOSF Evaluation process is shown in Figure 5.
4.5.3 Generate
a list of impacted SUPPO RT/ SUPPORTED Systems.
4.5.3.1 Considering the Conditi ons identified in step 4.5.2 as well any LCO Condition(s) previously in effect, determine if required SUPPORT or SUPPORTED SYSTEM(s) are rendered inoperable on redundant safety-related trains.
Train A Train B System i System i
System ii System ii Inoperable system System iii System iii System iv System iv
For the above example, IF Train A System iii is inoperable THEN Train B Systems i, ii and iii (support systems) and System iv (supported system) MUST be verified operable.
IF all Conditions in effect are limited to a single train, THEN no LOSF exists. All applicable Conditions SHOULD be entered, t he provisions of LCO 3.0.6 may be applied , and NO additional evaluation is required.
Below is an example of a list when Un it 2 SSPS is rendered inoperable while performing U2 RTB testing.
Approved By J.B. Stanley Vogtle Electric Generating Plant Procedure Version 10008-C 30 Effective Date 02/08/2013 RECORDING LIMITING CO NDITIONS FOR OPERATION Page Number 16 of 32 Printed October 3, 2013 at 12:42
4.5.4 Procedure
10005-C SHALL be used to manually ILLUMINATE SSMP for the systems/components identified in steps 4.5.2 and 4.5.3.
Approved By J.B. Stanley Vogtle Electric Generating Plant Procedure Version 10008-C 30 Effective Date 02/08/2013 RECORDING LIMITING CO NDITIONS FOR OPERATION Page Number 17 of 32 Printed October 3, 2013 at 12:42
4.5.5 Using
flowchart (Figure 5) determine if a Loss of Safety Function will exist IF the component/system is rendered inoper able. A method of place keeping should be used ensuring correct flow path is used. A (SRO) SHALL conduct an independent peer check of flowchart.
4.5.6 Determine
IF concurrent inoperable SUPPORT or SUPPORTED systems on required redundant train, results in the loss of a credited safety function.
4.5.6.1 Equipment supported by an inoperable Offsite Source OR Diesel Generator should NOT be considered inoperable for the purpose of this evaluation, UNLESS required by LCO 3.8.1 Required Ac tion A.2 or B.3. IF LCO 3.8.1 Condition A OR Condition B is in effect AND implementation of Required Action A.2 or B.3 subsequently results in t he inoperability of a required supported system, THEN a LOSF Evaluation MUST be re-performed.
Approved By J.B. Stanley Vogtle Electric Generating Plant Procedure Version 10008-C 30 Effective Date 02/08/2013 RECORDING LIMITING CO NDITIONS FOR OPERATION Page Number 18 of 32 Printed October 3, 2013 at 12:42 4.5.6.2 The TS related systems that SHOULD be evaluated when determining if a potential loss of safety function exists are:
Reactor Trip System Automatic Trip Logic Reactor Trip and Bypass Breakers ESFAS Automatic Actuation Logic & Actuation Relays: Safety Injection Containment Spray Containment Isolation Steamline Isolation Turbine Trip and Feedwater Isolation Auxiliary Feedwater Containment Sump Semi-automatic Switchover LOSP Instrumentation - Loss of either Undervoltage or Degraded Voltage Functions CVI Automatic Actuation Logic & Actuation Relays CREFS Automatic Actuation Logic & Actuation Relays High Flux at Shutdown Alarm (HFASA) Decay Heat Removal (including refueling operations) Pressurizer PORVs and associated Block Valves Cold Overpressure Protection System ECCS (See Step 4.5.6.8) Containment Penetrations Containment Spray and Cooling Systems Main Steam Isolation Valves MFIVs, MFRVs, and associated Bypass Valves Atmospheric Relief Valves Auxiliary Feedwater System Component Cooling Water System Nuclear Service Cooling Water System Ultimate Heat Sink Control Room Emergency Filtration Systems Piping Penetration Area Filtration and Exhaust System ESF Room Cooler and Safety-Related Chiller System AC Sources (including Safety Systems Sequencer) Diesel Fuel Oil, Lube Oil, Starting Air, and Ventilation DC Sources Inverters Electrical Distribution Systems
Approved By J.B. Stanley Vogtle Electric Generating Plant Procedure Version 10008-C 30 Effective Date 02/08/2013 RECORDING LIMITING CO NDITIONS FOR OPERATION Page Number 19 of 32 Printed October 3, 2013 at 12:42 4.5.6.3 A credited safety function is a func tion required to mitigate the consequences of a design basis event as described in the FSAR (reference FSAR Chapters 6
and 15), including all assumptions of the initiating event such as loss of offsite
power. NOTE FSAR assumptions such as loss of offsite power are considered as part of the initiating event and should not be considered as an "additional concurrent failure" in the following step.
4.5.6.4 A LOSF exists WHEN; assuming t hat with no additional concurrent failure during a design basis event, a required safety function assumed in the accident analysis CANNOT be performed.
4.5.6.5 If a LOSF is determined to exis t, the appropriate Conditions and Required Actions of the LCO in which the LOSF exists SHALL be entered. IF no Condition within the LCO addresses the LOSF, THEN LCO 3.0.3 shall be entered. Results of the LO SF Evaluation SHOULD be entered in the Unit Control Log and/or by initiation of an LCO tracking sheet documenting the LCO in which the LOSF exists.
4.5.6.6 IF a LOSF does not exist, THEN the Required Actions for the LCO SUPPORT SYSTEM(s) address the condition AND the required actions of the SUPPORTED SYSTEM(s) do NOT have to be performed as permitted by LCO 3.0.6. 4.5.6.7 Ensure that the Completion Time of any SUPPORTED SYSTEM has not been inappropriately extended as shown in Fi gure 6. Completion Time Extensions are considered inappropriate if the SUPPORTED SYSTEM remains inoperable for longer than the allowed out of serv ice time of the SUPPORT SYSTEM which caused the initial inoperability.
4.5.6.8 Technical Specific ation 3.5.2 Condition A, allows an ECCS train to be inoperable for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provided that at least 100% of the ECCS flow
equivalent to a single OPERABLE ECCS Train remains available. Analyses
have been performed for many of the potentia l flowpaths available that can be used to credit this allowance (reference 5.3). For cases where it is unclear if the
flow equivalent of a single ECCS tr ain remains operable, system engineering should be contacted for guidance.
Approved By J.B. Stanley Vogtle Electric Generating Plant Procedure Version 10008-C 30 Effective Date 02/08/2013 RECORDING LIMITING CO NDITIONS FOR OPERATION Page Number 28 of 32 Printed October 3, 2013 at 12:42 LOSF EVALUATION FLOWCHART Figure 5
- 1. 076AG2.4.47 001/LOIT AND LOCT/SRO/C/A 4.2/4.2/076G2.4.47/LO-TA-63013///
Initial condition:
- Unit 1 is at 100% reactor power.
Current conditions:
- 18013-C, "Rapid Power Reduction," is entered du e to a secondary transient.
- The foll owing reactor power trends are recorded:
TIME POWER 1100 100%
1115 97%
1130 93%
1145 89%
1200 86%
1215 81%
1230 79%
Which one of the following completes the following statement?
Chemistry sampling of the RCS __(1)__ required per Te ch Spec 3.4.16, "RCS Specific Activity," Surveillance Requirements, and per the Bases of Tech Spec 3.4.16, "RCS Specific Activity," the required action to reduce RCS Tavg below 500°F if the gross specific activity is exceeded is to prevent opening of the __(2)__.
__(1)__ __(2)__ is Atmospheric Relief Valves is Main Steam Safety Valves is NOT Atmospheric Relief Valves
is NOT Main Steam Safety Valves A.B.C.D.K/A 076 High Reactor Coolant ActivityG2.4.47 Ability to diagnose and recognize trends in an accurate and timelymanner utilizing the appropriate control room reference material.K/A MATCH ANALYSISMonday, February 24, 2014 4:10:27 PM 1
The question sets up a plausible scenario which includes all the required KA elements.
First the SRO candidate must recognize in a timely manner the requirements for RCS sampling following the power r eduction trend provided in the stem. Timely RCS activity sampling following a power reduc tion of more than 15% in o ne hour is required verify no fuel damage, which leads to high coolant activity. Then the candidate must determine the Tech Spec bases for loweri ng the energy level in the RCS if limits are exceeded. EXPLANATION OF REQUIRED KNOWLEDGE TS SR 3.4.16.2 requires veri fication of DOES EQ UIVALENT I-131 specific activity less than or equal to 1.0 uCi/gm between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> follo wing a power change of greater than or equal to 15% RTP within a 1 hr period. By definiti on, the 1 hr period is a rolling hour. As stated in the stem , the power change between 1100 and 1200 is 14%RTP.
The power change between 1115 a nd 1215 is 18%. Therefore, the 15% RTP in an hour has been exceeded.
Per TS 3.4.13, if I-131 is in excess of limits or LCO co mpletion time of Cond A or B cannot be met, then the plant is required to be placed in Mode 3 with RCS Tavg <500F.
Per TS 3.4.13 Bases, with RCS Tavg <500F, the release of radioactivity in the event of a SGTR is unlikely since the saturation pressure of the reac tor coolant is below the lift pressure settings of the main steam safety valve lift setpoint.ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. Pa rt 1 is correct the candidate should deter mine that from the stem information provided that a power reduction of more than 15% in one hour occured bet ween 1115 and 1215 and require chemistry sampling of the RCS to verify no fuel damage.
Part 2 is incorrect however 'pl ausible' since the candidate may determine that the bases for t he RCS temperature limit is to prevent the ARVs from lifting si nce they would normally open an a lower setpoint.B. Correct. Part 1 is correct.
See Part 1 of choice A above.Part 2 is correct. Per Tech Sp ec 3.4.16 Bases, the purpose of lowering RCS temperature below 500°F is to prevent radioactive releases due to main stream safety valves lifting.
C. Incorrect. Plausible. Part 1 is incorr ect however 'plausible' since the candidate may determine based on the time li ne given that RCS sampling is not required. The rolling hour starting at times 1100 and 1130 are <15% change.
Part 2 is incorrect. See Part 2 of choice A above.
D. Incorrect. Plausible. Part 1 is in correct. See Part 1 of choice C above.
Part 2 is correct. See Part 2 of choice B above.Monday, February 24, 2014 4:10:27 PM 2
SRO JUSTIFICATION(2) Facility operating limitations in the technical specifications and their bases.-Can question be answered solely by knowing = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action?
No, the question requires Bases knowledge.-Can question be answered solely by knowing the LCO/TRM information listed "above-the-line?"
No, the applicability statement for TS 3.4.
16 does state Mode 3>500F Tavg. However, the Bases of this applicability and the Required Actionsfor both Conditon B & C are only listed in the Bases document.
-Can question be answered solely by knowing the TS Safety Limits? No, SafetyLimits are not addressed by this question.-Does the question involve one or more of the following for TS,TRM, or ODCM?
- Application of Required Actions (Sec tion 3) and Surve illance Requirements (Section 4) in accordance with rules of applicati on requirements (Section 1)* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)* Knowledge of TS bases that is required to analyze TS required actions and terminology Yes, the questio n requires the candida te to specificallyknow the Bases for TS 3.4.13.
Level: SROTier # / Group # T1 / G2 K/A# 076G2.4.47 Importance Rating: 4.2 / 4.2Technical
Reference:
TS 3.4.16, Ammendment No. 158, page 4.3.16-2 TS Bases 3.4.16, Rev 1-10/01, page B 3.4.16-3References provided: None Learning Objective: LO-TA-63013 Implement Techni cal Specification LCO using 10008-C (SRO Only)
LO-PP-16001-04 St ate the LCO, applicability, bases, and the 1 hr or less actions for each of the following: 3.4.16 RCS Specific ActivityQuestion origin: NEW Cognitive Level: C/A 10 CFR Part 55 Content: 41.5 / 43.2 Comments: You have completed the test!Monday, February 24, 2014 4:10:27 PM 3
RCS Specific Activity 3.4.16 Vogtle Units 1 and 2 3.4.16-2 Amendment No. 158 (Unit 1)
Amendment No. 140 (Unit 2)
ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and associated Completion
Time of Condition A not
met. OR DOSE EQUIVALENT I-131 in the
unacceptable region of
Figure 3.4.16-1.
C.1 Be in MODE 3 with Tavg < 500°F. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 Verify reactor coolant gross specific activity 100/ µCi/gm. In accordance with the Surveillance Frequency Control Program SR 3.4.16.2 ----------------------------NOTE-----------------------------
Only required to be performed in MODE 1.
Verify reactor coolant DOSE EQUIVALENT I-131 specific activity 1.0 µCi/gm.
In accordance with the Surveillance Frequency Control Program AND Between 2 and
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a
THERMAL POWER change of 15% RTP within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period
(continued)
RCS specific activity satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii). LCO The specific iodine activity is limited to 1.0 Ci/gm DOSE EQUIVALENT I-131, and the gross specific activity in the reactor coolant is limited to the number of Ci/gm equal to 100 divided by (average disintegration energy of the sum of the average beta and gamma energies of the coolant nuclides). The limit on DOSE EQUIVALENT I-131 ensures the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose to an individual at the exclusion area boundary during the Design Basis Accident (DBA) will be a small fraction of the allowed thyroid dose. The limit on gross specific activity ensures the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> whole body dose to an individual at the exclusion area boundary during the DBA will be a small fraction of the allowed whole body dose. The SGTR accident analysis (Ref. 2) shows that the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> site boundary dose levels are within acceptable limits. Violation of the LCO may result in reactor coolant radioactivity levels that could, in the event of an SGTR, lead to exclusion area boundary doses that exceed the 10 CFR 100 dose guideline limits. APPLICABILITY In MODES 1 and 2, and in MODE 3 with RCS average temperature 500F, operation within the LCO limits for DOSE EQUIVALENT I-131 and gross specific activity are necessary to contain the potential consequences of an SGTR to within the acceptable site boundary dose values. For operation in MODE 3 with RCS average temperature < 500F, and in MODES 4 and 5, the release of radioactivity in the event of a SGTR is unlikely since the saturation pressure of the reactor coolant is below the lift pressure settings of the main steam safety valves. (continued)
- 1. G2.1.34 001/LOIT AND LOCT/SRO/M/F 2.7/3.5/G2.1.34/LO-TA-60024/// Given the following:
- Unit 1 is at 100% reactor power.
Which one of the following completes the following statement?
Per Tech Spec 3.7.16, "Secondary Specific Ac tivity," the specific activity of the secondary coolant shall be
< __(1)__ Ci/gm Dose Equivalent I-131,and operating within this limit ensures that the off-site dose will be limited to within a small fraction of the 10 CFR 100 dose guideline values in the event of a __(2)__. __(1)__ __(2)__ 0.10 steam line break 0.10 loss of all AC power 1.0 steam line break 1.0 loss of all AC power A.B.C.D.Tuesday, February 25, 2014 9:35:23 AM 1
K/AG2.1.34 Knowledge of primary and secondary plant chemistry limits.K/A MATCH ANALYSIS The question tests the candidate's knowledge of primary and second ary plant chemistry limits by requiring the student to select Tech Spec 3.7.16, "S econdary Specific Activity," specific activity limit for Dose Equivalent I-131 and the a ssociated limiting design bases accident. This value is pitte d against the limit for Tech Spec 3.4.16, "RCS Specific Activity".EXPLANATION OF REQUIRED KNOWLEDGE Per Tech Spec 3.7.16, "Secondary Specific Ac tivity" bases, the acci dent analysis of the main steam line break (MSLB), assumes the initial secondar y coolant specific activity to have a radioactive isotope concentration of 0.10 µCi/gm DOSE EQUIVALENT I-131.
This assumption is used in the analysis for determining the radiological consequences of the postulated accident. T he accident analysis, based on this and other assumptions, shows that the radiological consequences of an MSLB do not exceed a small fraction of the unit EAB limits for whole body and thyroid dose rates.
Per Tech Spec 3.4.16, "RCS Specific Activity" bases, the limit of 1.0 µCi/gm DOSE EQUIVALENT I-131 on specific activity of ensures that th e doses are held to a small fraction of the 10 CR F 100 limits during a st eam generator tube ru pture accident. The maximum dose to the whole body and the thyroid considers ex posure to an individual at the exclusion boun dary for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.ANSWER / DISTRACTOR ANALYSISA. Correct. Part 1 is correct. Per Tech Spec 3.7.16, 'Secondary Specific Activity', the specific activity of the secondary co olant shall be <
or = 0.10 µCi/gm DOSE EQUIVALENT I-131.
Part 2 is correct. Per Tech Spec 3.7.16, 'Secondary Specific Activity' bases, the accident anal ysis of the main steam line break (MSLB), assumes the init ial secondary coolant specific activity to have a radioactive isotope concentration of 0.10
µCi/gm DOSE EQUIVALENT I-131.
This assumption is used in the analysis for determining the radiological consequences of the postulated accident. The a ccident analysis, based on this and other assumptions, show s that the radiological consequences of an MSLB do not exceed a small fraction of the unit EAB limits for whole body and thyroid dose rates.B. Incorrect. Plausible. Part 1 is correct. See Part 1 of choice A above.
Part 2 is incorrect however 'pl ausible' since the candidate mayTuesday, February 25, 2014 9:36:10 AM 1
recall the LOSP as the most limiting accident not taking into account the rema ining steam generators are available for core decay heat dissipation by vent ing steam to the atmosphere through the MSSVs and steam ge nerator atmospheric dumpvalves (ARVs). The Auxiliary Fe edwater System supplies the necessary makeup to the steam g enerators. Vent ing continues until the reactor coolant te mperature and pressure have decreased sufficiently for the Resi dual Heat Removal System to complete the cooldown.
C. Incorrect. Plausible. Part 1 is incorr ect however 'plausible' since it's reasonable to assume the candidate may confus e primary limit for specific activity with the secondary lim it since the numbers are only distinguish by the movem ent of the decimal point. Part 2 is correct. See Part 2 of choice A above.
D. Incorrect. Plausible. Part 1 is in correct. See Part 1 of choice C above.
Part 2 is correct. See Part 2 of choice B above.SRO JUSTIFICATION (10CFR43(b))(2) Facility operating limitations in the technical specifications and their bases.-Can question be answered solely by knowing = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action?
No, thequestion is not addressing Tech Spec action times.-Can question be answered solely by knowing the LCO/TRM information listed "above-the-line?"
No, the question is not addr essing Tech Spec above-the-lineinformation.
-Can question be answered solely by knowing the TS Safety Limits?
No, thequestion is not related to Tech Spec Safety Limits.-Does the question involve one or more of the following for TS,TRM, or ODCM?
- Application of Required Actions (Sec tion 3) and Surve illance Requirements (Section 4) in accordance with rules of applicati on requirements (Section 1)* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)* Knowledge of TS bases that is required to analyze TS required actionsand terminology.
Yes, the answer to the question is only found in Tech Spec bases.Tuesday, February 25, 2014 9:36:10 AM 2
Level: SROTier # / Group # T3 K/A# G2.1.34 Importance Rating: 2.7 / 3.5Technical
Reference:
TS 3.4.16, Ammendment No. 137, page 3.4.16-1 TS 3.7.16, Ammendment No. 158, page 3.7.16-1 TS Bases 3.4.16, Re v 0, page B 3.4.16-1 TS Bases 3.7.16, Rev 1-10/01, page B 3.7.16-2References provided: None Learning Objective: LO-TA-63013 Implement Techni cal Specification LCO using 10008-C (SRO Only)
LO-PP-16001-04 St ate the LCO, applicability, bases, and the 1 hr or less actions for each of the following: 3.4.16 RCS Specific Activity.
LO-LP-39211-01 For any given it em in section 3.7 of Tech Specs, be able to:
- a. State the LCO.
- b. State any one ho ur or less required actions.LO-TA-60024 Respond to Abnormal Plant Chemistry per 18014-C or 18015-CQuestion origin: NEW Cognitive Level: M/F
10 CFR Part 55 Content: 41.5 / 41.10 / 43.2Comments: You have completed the test!Tuesday, February 25, 2014 9:36:10 AM 3
RCS Specific Activity 3.4.16 Vogtle Units 1 and 2 3.4.16-1 Amendment No. 137 (Unit 1)
Amendment No. 116 (Unit 2)
3.4 REACTOR
COOLANT SYSTEM (RCS)
3.4.16 RCS Specific Activity
LCO 3.4.16 The specific activity of the reactor coolant shall be within limits.
APPLICABILITY: MODES 1 and 2, MODE 3 with RCS average temperature (Tavg) 500°F.
ACTIONS
NOTE--------------------------------------------------------
LCO 3.0.4c is applicable.
CONDITION REQUIRED ACTION COMPLETION TIME A. DOSE EQUIVALENT I-131 > 1.0
µCi/gm. A.1 Verify DOSE EQUIVALENT I-131
within the acceptable
region of Figure 3.4.16-1.
AND A.2 Restore DOSE EQUIVALENT I-131 to
within limit.
Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> B. Gross specific activity of the reactor coolant not
within limit.
B.1 Perform SR 3.4.16.2.
AND B.2 Be in MODE 3 with Tavg < 500°F. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (continued)
Secondary Specific Activity 3.7.16 3.7 PLANT SYSTEMS 3.7.16 Secondary Specific Activity
LCO 3.7.16 The specific activity of the secondary coolant shall be 0.10 µCi/gm DOSE EQUIVALENT I-131.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Specific activity not within limit.
A.1 Be in MODE 3.
AND A.2 Be in MODE 5.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.16.1 Verify the specific activity of the secondary coolant is 0.10 µCi/gm DOSE EQUIVALENT I-131. In accordance with the Surveillance Frequency Control Program RCS Specific Activity B 3.4.16 Vogtle Units 1 and 2 B 3.4.16-1 Revision No. 0 B 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4.16 RCS Specific Activity BASESBACKGROUND The maximum dose to the whole body and the thyroid that an individual at the exclusion area boundary can receive for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during an accident is specified in 10 CFR 100 (Ref. 1). The limits on specific activity ensure that the doses are held to a small fraction of the 10 CFR 100 limits during analyzed transients and accidents. The RCS specific activity LCO limits the allowable concentration level of radionuclides in the reactor coolant. The LCO limits are established to minimize the offsite radioactivity dose consequences in the event of a steam generator tube rupture (SGTR) accident. The LCO limits specific activity for both DOSE EQUIVALENT I-131 and gross specific activity. The allowable levels are intended to limit the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dose at the exclusion area boundary to a small fraction of the 10 CFR 100 dose guideline limits. The limits in the LCO are standardized, based on parametric evaluations of offsite radioactivity dose consequences for typical site locations. The parametric evaluations showed the potential offsite dose levels for a SGTR accident were an appropriately small fraction of the 10 CFR 100 dose guideline limits. Each evaluation assumes a broad range of site applicable atmospheric dispersion factors in a parametric evaluation. APPLICABLE The limits on the specific activity of the reactor coolant SAFETY ANALYSES ensures that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed a small fraction of the 10 CFR 100 dose guideline limits following a SGTR accident. The SGTR safety analysis (Ref. 2) assumes that the reactor has been operating at the maximum allowable Technical Specification limit for primary coolant activity and primary to secondary leakage for sufficient time to establish equilibrium concentrations of radionuclides in the reactor coolant and in the secondary coolant. (continued)
Secondary Specific Activity B 3.7.16 Vogtle Units 1 and 2 B 3.7.16-2 Rev. 1-10/01 BASES (continued) APPLICABLE The accident analysis of the main steam line break (MSLB), SAFETY ANALYSES as discussed in the FSAR, Chapter 15 (Ref. 2) assumes the initial secondary coolant specific activity to have a radioactive isotope concentration of 0.10 Ci/gm DOSE EQUIVALENT I-131. This assumption is used in the analysis for determining the radiological consequences of the postulated accident. The accident analysis, based on this and other assumptions, shows that the radiological consequences of an MSLB do not exceed a small fraction of the unit EAB limits (Ref. 1) for whole body and thyroid dose rates. With the loss of offsite power, the remaining steam generators are available for core decay heat dissipation by venting steam to the atmosphere through the MSSVs and steam generator atmospheric dump valves (ARVs). The Auxiliary Feedwater System supplies the necessary makeup to the steam generators.
Venting continues until the reactor coolant temperature and pressure have decreased sufficiently for the Residual Heat Removal System to complete the cooldown. In the evaluation of the radiological consequences of this accident, the activity released from the steam generator connected to the failed steam line is assumed to be released directly to the environment. The unaffected steam generator is assumed to discharge steam and any entrained activity through the MSSVs and ARVs during the event. Since no credit is taken in the analysis for activity plateout or retention, the resultant radiological consequences represent a conservative estimate of the potential integrated dose due to the postulated steam line failure. Secondary specific activity limits satisfy Criterion 2 of 10 CFR 50.36 (c)(2)(ii). LCO As indicated in the Applicable Safety Analyses, the specific activity of the secondary coolant is required to be 0.10 Ci/gm DOSE EQUIVALENT I-131 to limit the radiological consequences of a Design Basis Accident (DBA) to a small fraction of the required limit (Ref. 1). Monitoring the specific activity of the secondary coolant ensures that when secondary specific activity limits are exceeded, appropriate actions are taken in a timely manner (continued)
- 1. G2.1.37 001/LOIT AND LOCT/SRO/M/F 4.3/4.6/G2.1.37/LO-LP-63500-09///
Given the following: - Unit 1 is at 3% reactor power and raising power following an outage.
- Unit 2 is at 100% reactor power.
- The Shift Manager and the following assigned personnel are in the control room:
Unit 1 Unit 2 Shift Supervisor Shift Supervisor Reactivity Management SRO 1 NPO*
2 NPOs*(*) NPO - Individual with a Reactor Operator License.
Which one of the following completes the following statement?
Per TRM 15.1, "Unit Staffing," the total number of NPOs assig ned __(1)__ meet the minimum required for the given conditions, and per NMP-OS-001, "Reactivity Management Program," an y changes to the Unit 1reactivity plan must be approved by the __(2)__. __(1)__ __(2)__ does Reactivity Management SRO does Shift Supervisor does NOT Reactivity Management SRO does NOT Shift Supervisor A.B.C.D.K/A Conduct of Operations 2.1.37 Knowledge of procedures, guidelines, or limitations associated withreactivity management.K/A MATCH ANALYSIS The question tests the candidat e's knwoledge of administrati ve procedural requirements assoicated with the required approvals for changing reactivity plans during low power ascention following a reactor startup.Tuesday, February 25, 2014 9:43:40 AM 1
EXPLANATION OF REQUIRED KNOWLEDGE The candidate is required to evaluate the mini mum shift staffing for the given Modes per TRM 15.1 for the RO positions. Per Table 15.1.2-1, (3)
ROs are required with both Units in Modes 1-4. (2) RO's must be a ssigned to the OATC posit ion on each of the Units.Per NMP-OS-001 step 6.1.2.1, the Shift M anager, who has ultimate responsibility for controlling the reactor core, approves formal Reactivity Management Plans as described in paragraph 6.6. Th e Shift Supervisor, or a designated Senior Reactor Operator, directly supervises r eactivity changes. A written plan is developed by reactor engineering and approved for significant reacti vity changes. The Shift Manager, Shift Supervisor, OATC, Shift Technical Advisor, and Reactor Engineer co ncur on Reactivity Management Plans and changes ther eto, prior to implementation.ANSWER / DISTRACTOR ANALYSISA. Incorrect. Plausible. The first part is correct. Per TRM 15.1 Table 15.1.2-1, (3) ROs are required with both Units in Modes 1-4. (2) RO's must be assigned to the OATC position on each of the Units. This condition is met.
The second part is incorrect.
Per NMP-OS-001 step 6.1.2.1, the Shift Manager, Shift Supervisor, OATC, Shift Technical Advisor, and Reactor Engi neer concur on Reactivity Management Plans and c hanges thereto, prior to implementation. However, t he Reactivity Management SRO has oversight of all reactivity changes. As such, it is reasonable for a candidate without specific knowledge of the procedural requirements to assume t he Reactivity Management SRO would also have this authority.
Therefore, this distractor is plausible.B. Correct. The first part is correct. See the first part of choice A above.
The second part is correct. Per NMP-OS-001 step 6.1.2.1, the Shift Manager, Shift S upervisor, OATC, Shif t Technical Advisor, and Reactor Engineer concur on Reactivity Management Plans and changes thereto, prio r to implementation.
C. Incorrect. Plausible. The first part is incorrect. Per TRM 15.1 Table 15.1.2-1, (3)
ROs are required with bot h Units in Modes 1-
- 4. (2) RO's must be assigned to the OATC position on each of the Units. This condition is met. However, administrative procedure 00012-C
'Shift Manning' limit s are not currently me
- t. Per 00012-C, the ENN communication position is norma lly filled by each of the UOs which results in (4) NPOs being required.
This is the normal shift alignment. It is reasonable for a candidate who does not have adequat e knowledge of TRM 15.1 requirementsto determine insufficient NPOs ex ist. Therefore, this distractor is plausible.Tuesday, February 25, 2014 9:43:40 AM 2
The second part is inco rrect. See the second part of choice A above.D. Incorrect. Plausible. The first part is in correct. See the first part of choice C above.
The second part is correct. See the second part of choice B above.SRO JUSTIFICATION (10CFR43(b))(5) Assessment of facility conditions and selection of appropriate proceduresduring normal, abnormal, and emergency situations.-Can the question be answered solely by knowing "systems knowledge", i.e.,how the system works, flowpath, logic, component location?
No, the questionis not system related in any way.-Can the question be answered solely by knowing immediate operator actions? No, the question does not involve any operator actions.-Can the question be answered solely by knowing entry conditions for AOPs orplant parameters that require direct entry to major EOPs?
No, the question does not involve an AOP or EOP.-Can the question be answered solely by knowing the purpose, overall sequenceof events, or overall mitigative strategy of a procedure?
No, the questionrequires the knowledge of a specifc detail. Overall understanding of approvalsactually directs the candidate to an incorrect answer.-Does the question re quire one or more of the following?
- Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or se ction of a procedure to mi tigate, recover, or with which to proceed
- Knowledge of when to im plement attachments and appendices, including how to coordinate these it ems with procedure steps
- Knowledge of diagnostic st eps and decision points in the EOPs that involve transitions to event specific s ubprocedures or emergency contingency procedures* Knowledge of administrative procedures that specify hierarchy,implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
Yes, the question requires specific knowledge ofthe hierachy of an administrative procedure for approving Reactivity Planchanges once developed.(6) Procedures and limitations involved in initial core loading, alternations incore configuration, control rod programming, and determination of various internal and external effects on core reactivity. Yes, the question requires the candidate to have specific knowledge of theapproval process and responsibilities of control room personnel that approveReactivity Plans and overs ee Reactivity Manipulations.Tuesday, February 25, 2014 9:43:40 AM 3
Level: SROTier # / Group # T3 K/A# G2.1.37 Importance Rating: 4.3 / 4.6Technical
Reference:
NMP-OS-001 Rev 17.0, page 9 00012-C Rev 17.2, pages 5&6
TRM 15.1, Table 15.1.2-1 Rev 0 12/26/96, page 15.1-2References provided: None Learning Objective: LO-LP-36110-01 Per Technical Specific ations, state the shift manning requirements. (SRO)
LO-LP-63503-02 State the requ irements of the OATC with regards to shift m anning when fuel is in either reactor.
LO-LP-63503-05 State the requ irements of shift manning for the following conditions: (SRO only)f. minimum shift crew LO-LP-63510-04 Name the two site gr oups that have the most day-to-day effect on reactivity management.
LO-LP-63500-09 State reactivity mani pulation expectations including: monitoring, briefing, pre-plans,peer checks, transient operation, pull-and-wait, load change concurrence, instrument res ponse, changing temperature and power, and Reactivity
Management SRO.
Question origin:
BANK - Hatch 2011 NRC # G2.1.37Cognitive Level: M/F 10 CFR Part 55 Content: 43.5 / 43.6 Comments: You have completed the test!Tuesday, February 25, 2014 9:43:40 AM 4
Table 15.1.2-1 MINIMUM SHIFT CREW COMPOSITION TWO UNITS WITH A COMMON CONTROL ROOM Unit Staff TR 15.1 Position Number of Individuals Required to Fill Position Both Units in MODE 1, 2, 3, or 4 ss 1 SRO 1 RO 3(2) NLO 3(2) STA 1 (3) Both Units in MODE 5 or 6 or DEFUELED 1 None(1 l None One Unit in MODE 1, 2, 3, or4 and One Unit in MODE 5 or 6 or DE FUELED 1 1 (1) At least one licensed Senior Operator or licensed Senior Operator Limited to Fuel Handling who has no other concurrent responsibilities must be present during CORE ALTERATIONS on either unit. (2) (3) STA-At least one of the required individuals must be assigned to the designated position for each unit. See TS 5.2.2.g. Shift Superintendent with a Senior Operator License. Individual with a Senior Operator License. Individual with an Operator License. Non-licensed operator.
Shift Technical Advisor. Vogtle Units 1 and 2 Technical Requirement 15.1 -2 Rev. 0 12/26/96 Approved By J. B. Stanley Vogtle Electric Generating Plant Procedure Number Rev 00012-C 17.2 Date Approved 03/17/2009 SHIFT MANNING REQUIREMENTS Page Number 5 of 6 Printed November 22, 2013 at 16:32 DATA SHEET 1 Sheet 1 of 2 Minimum Shift Manning (Either Unit in Mode 1
-4) Date: Shift (Day/Night): POSITION UNIT #1 COMMON UNIT #2 Shift Manager V-OPS-SS, V-ERO-CR01, and V-ERO-CR10 Also assigned as Emergency Director SS V-OPS-USS, V-ERO-CR02, AND V-ERO-CR10 Also assigned as ENS Communicator Also assigned as ENS Communicator OATC V-OPS-RO/BOP UO V-OPS-RO/BOP and V-ERO-CR04 Also assigned as ENN Communicator Also assigned as ENN Communicator SO V-OPS-SO SO/NPO SO/NPO STA (May be assigned other duties)
V-OPS-STA (SM, or SSS or SS not assigned to FB or ENN Communicator)
Fire Team Captain V-FP-FIRE BRIGAD E LEADER SSS, or SS C&T FB Member
FB Member
- 4. V-FP-FIRE BRIGADE SO Security V-ERO-SEC or V-ERO-SEC02 Per Security Procedure 90101-C SAT Operator
- 5. V-OPS-SO-OAO Assigned per procedure 13419-C SO/NPO/SRO Wilson Operator
- 6. V-OPS-WILSON BLKSTRT Assigned per procedure 13419-C SO/NPO Emergency Plan POSITION UNIT #1 COMMON UNIT #2 Emergency Director V-OPS-SS Shift Manager ENN Communicator V-ERO-CR04 or V-ERO-CR10 UO Unaffected Unit UO Unaffected Unit ENS Communicator V-OPS-USS or V-OPS-STA SS Unaffected Unit SS Unaffected Unit
Approved By J. B. Stanley Vogtle Electric Generating Plant Procedure Number Rev 00012-C 17.2 Date Approved 03/17/2009 SHIFT MANNING REQUIREMENTS Page Number 6 of 6 Printed November 22, 2013 at 16:32 DATA SHEET 1 Sheet 2 of 2 Emergency Plan (cont) POSITION UNIT #1 COMMON UNIT #2 Dose Assessment V-ERO-TSC09 OR V-ERO-TSC10 HP Foreman Field Monitoring Team (FMT)
- 2. HP Tech/Chem Tech/SO/I&C Tech FMT Communicator V-ERO-TSC18 Chem Foreman/Chem Tech/HP Tech/ Maint. Shift ATL Chemistry Sampler V-ERO-OSC09 Chemistry Tech Mechanical Repair and Corrective Action V-ERO-OSC07 Mechanic Electrical Repair and Corrective Action V-ERO-OSC06 Electrician I & C Repair and Corrective Action V-ERO-OSC08 I & C Technician In Plant Monitors
- 1. V-ERO-OSC17 HP Technician
- 2. HP Technician Search & Rescue/First Aid
- 1. (May be assigned other Duties) V-ERO-OSC15 HP Technician
- 2. HP Technician Minimum Dual Unit Safe Shutdown POSITION UNIT #1 COMMON UNIT #2 Emergency Director SM ENN UO Unit #1 or #2 ENS UO Unit #2 or #1 Shutdown Panel "B" SS SS Shutdown Panel "A" OATC OATC Shutdown Panel "C" SO SO Fire Brigade Same as Normal Operations
- 1. Personnel may NOT be assigned to more than one position unless specifically noted next to the position label.
- 2. If both units are in Modes 5, 6, or defueled, minimum shift manning for operations may be r educed per Operations Manager (not Emergency Plan or Fire Brigade staffing).
COMMENTS:
Approved by: Date: Time: Shift Manager
Southern Nuclear Operating Company Nuclear Management Procedure Reactivity Management Program NMP-OS-001 Version 17.0 Page 9 of 39
6.1.2 Expectations
6.1.2.1 Reactivity Management Controls NOTE: Prior to implementing any activity that has the potential to add positive reactivity; the plant shall be ramped down, as necessary, to ensure that the reactor does not exceed 100.0% of the unit's Rated Thermal Power limit.
Reactivity management involves a systematic process of controlling evolutions with the potential to affect reactivity:
Planned reactivity changes are conducted in a controlled and conservative manner Unexpected changes in reactivity are minimized Conservative actions are taken in response to unexpected reactivity changes Reactivity control systems, including r eactivity monitoring instrumentation, are available and reliable Modifications, analyses, and predictions are correct and effectively implemented The Shift Manager, who has ultimate responsibility for controlling the reactor core, approves formal Reactivity Management Plans as described in paragraph 6.6. The Shift Supervisor, or a designated Senior Reactor Operator, directly supervises reactivity changes. A strong relationship exists between reactor engineering and operations. Reactor engineering is actively engaged in the planning of significant reactivity changes such as reactor startup, reactor shutdown, planned power changes, and special tests with the potential to affect reactivity. A written plan is developed by reactor engineering and approved for significant reactivity changes. The Shift Manager, Shift Supervisor, OATC, Shift Technical Advisor, and Reactor Engineer concur on Reactivity Management Plans and changes thereto, prior to implementation.
6.1.2.2 Control Room Operations Only operators with an active license (NPO or SRO) manipulate the controls of the reactor. An individual in a NRC-approved license training program may manipulate reactor controls when under the direction and in the presence of a licensed operator.
The OATC obtains concurrence from the Shift Supervisor prior to allowing or performing planned reactivity manipulations. Directions that affect reactivity go through the Shift Supervisor.
A briefing is conducted prior to the start of a planned reactivity change. The reactor operator performing rod movement, the OATC and Shift Supervisor monitor reactivity manipulations and verify that the end state of the reactivity manipulation is as expected.
- 1. G2.2.20 001/LOCT AND LOIT/SRO/M/F 2.6/3.8/G2.2.20/LO-LP-63350-07//HL18 NRC/
Initial condition:
- ALB34-D0 1 125 VDC SWGR 1AD1 TROUBLE is received due to a bus ground.
Current conditions:
- Per NMP-AD-002, "'Problem Solving and Troubleshooting Guidelines," a troubleshooting plan has been written.
- As part of t he plan, operations personnel will open various breakers and maintenance personnel will open links to measure resistance inside the 1AD1 panel.Which one of the following completes the following statement?
This type of Troubleshooting Monitoring is called __(1)__, and the tracking of the equipment out-of-service time while troubleshooting 1AD1 is the responsibility of th e __(2)__ Department.(1) Intrusive (2) Maintenance(1) Intrusive (2) Operations (1) Non-Intrusive (2) Maintenance (1) Non-Intrusive (2) Operations A.B.C.D.G2.2.20 Equipment ControlKnowledge of the process for ma naging troubleshooting activities.K/A MATCH ANALYSIS:
The candidate is presented with a plausible scenario where Troubl eshooting is to be performed in the 1AD1 Panel by both maintenance personnel and operations personnel. The activity invo lves opening links, breakers and measuring resistance toTuesday, February 25, 2014 10:15:47 AM 1
resolve a ground problem. The candidate must determine if this is intrusive or non-intrusive trouble shootin g and also has to determine the person responsible formaintaining the system status of the panel during Trouble Shooting activities.EXPLANATION OF REQUIRED KNOWLEDGE Per NMP-AD-002 definition 3.
5, non-intrusive monitoring is defined as the act ofmonitoring a component or system by not affecting normal oepration of the conponent
or system.
Per NMP-AD-002 definition 3.6, intrusive monitoring is defined as the act of temporarilyaltering the system to allow monitoring a component or system. This applies to electrical or mechani cal testing methods.
Per NMP-AD-002 responsibility 4.2, the Operation Depar tment is responsible formaintaining approved system status during troubleshooting activities (ie Out of Service).ANSWER / DISTRACTOR ANALYSIS:
A. Incorrect. Plausible.
Part 1 is correct. Opening br eakers by Operations and opening links by Maintenance is an intrusive troubleshooting per
NMP-AD-002.
Part 2 is incorrect. Operations Department is responsible fortracking systems status, however the answer is 'plusable'
because opening links, measuring resistance, etc. would all be maintenance activities and so me amount of individual component status control is r equired. These components are typically tracked using a "lifted lead" sheet inside the MWO. Even though maintenance tracks status of some components, Operations still retains over all responsibility for ensuring equipment status and c onfiguration control.B. Correct. Part 1 is correct.
See Part 1 of choice A above.
Part 2 is correct. Per NMP-AD-002 responsibility 4.2, the Operation Department is respons ible for maintaining approvedsystem status during troubleshooting activities.C. Incorrect. Plausible. Part 1 is incorrect however 'plusable' because the word intrusive is very subjective and would require procedure knowledge to make this dist inction. The candidate may determine that allowing intrusiv e troubleshooting on 1AD1 with the unit on line would imply t oo much risk to operations andthus eliminate this possibility.
Part 2 is incorrect. See Part 2 of choice A above.
D. Incorrect. Plausible. Part 1 is in correct. See Part 1 of choice C above.
Part 2 is correct. See Part 2 of choice B above.Tuesday, February 25, 2014 10:15:47 AM 2
SRO JUSTIFICATION(5) Assessment of facility conditions and selection of appropriate proceduresduring normal, abnormal, and emergency situations.-Can the question be answered solely by knowing "systems knowledge", i.e.,how the system works, flowpath, logic, component location? No, systemknowledge will not answer any part of this question.-Can the question be answered solely by knowing immediate operator actions? No, IOAs are not addressed in any way in this question.-Can the question be answered solely by knowing entry conditions for AOPs orplant parameters that require direct entry to major EOPs? No, this question doesnot pertain to an EOP or AOP.
-Can the question be answered solely by knowing the purpose, overall sequenceof events, or overall mitigati ve strategy of a procedure?
No, the question pertains to specific guidance in NMP-AD-002, which cannot be answered bybroad knowledge of the procedure.-Does the question re quire one or more of the following?* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or se ction of a procedure to mi tigate, recover, or with which to proceed* Knowledge of when to impl ement attachments and appendices, including how to coordinate these it ems with procedure steps* Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific s ubprocedures or emergency contingency procedures* Knowledge of administrative procedures that specify hierarchy,implementation, and/or coordination of plant normal, abnormal, and emergency procedures Yes, the question requires specific knowledge of an administrative proced ure that specifies the hiearchy by which plant configuration is statused and configuration control is maintained duringnormal plant operation.Tuesday, February 25, 2014 10:15:47 AM 3
Level: SROTier # / Group # T3 K/A# G2.2.20 Importance Rating: 2.6 / 3.8Technical
Reference:
NMP-AD
-002, Rev 10.0, pages 4 & 5References provided: None Learning Objective: LO-LP-63350-07 Define the following terms:
- f. Trouble shooting LO-LP-63354-03 Describe the Shift Manager's responsibility concerning maintenance activities.Question origin: BANK -
HL18 NRC # G.2.2.20Cognitive Level: M/F
10 CFR Part 55 Content:
41.10 / 43.5 Comments:You have completed the test!Tuesday, February 25, 2014 10:15:47 AM 4
Southern Nuclear Operating Company Nuclear Management Procedure Problem Solving and Troubleshooting Guidelines NMP-AD-002 Version 10.0 Page 4 of 14
1.0 Purpose
The purpose of this procedure is to provide a process for performance of troubleshooting when required for plant problem resolution. These problems may include equipment failures, abnormal operating conditions, negative performance trends or recurring events.
2.0 Applicability
2.1 This procedure is applicable to troubleshooting activities at any of the SNC sites.
2.2 Entry
into the formal troubleshooting process is not intended for simple problems where the cause appears straightforward or known. In these cases investigation will be controlled by the Work Order process.
2.3 Formal
troubleshooting activities shall be performed in accordance with this procedure unless waived by plant management or management within the affected department. If waived, the justification shall be documented in the appropriate location (Condition Report, Work Order, etc).
2.4 This procedure does not apply to special tests.
2.5 All troubleshooting shall use high impedance M&TE and/or isolation transformers on the signal and AC power source unless low impedance is specifically called for in equipment procedure.
3.0 Definitions
3.1 Troubleshooting
- A systematic approach to data collection, failure analysis, or a measurement plan that results in high confidence that the complete cause of system/equipment degradation has been determined. There may be potential personnel safety risk.
3.2 High Risk Troubleshooting - Potential impacts are assessed as high risk when evaluated per NMP-DP-001, Operational Risk Awareness.
3.3 Medium
Risk Troubleshooting - Potential impacts are assessed as medium risk when evaluated per NMP-DP-001, Operational Risk Awareness.
3.4 Low Risk Troubleshooting - Potential impacts are assessed as low risk when evaluated per NMP-DP-001, Operational Risk Awareness.
3.5 Non-Intrusive Monitoring
- The act of monitoring a component or system by not affecting normal operation of the component or system. Examples would be using "Voltage Test Jacks," monitoring voltages across relay contacts, power supplies, etc.
3.6 Intrusive
Monitoring - The act of temporarily altering the system to allow monitoring a component or system. This applies to electrical or mechanical testing methods.
3.7 Stop-Decision Points - Administrative and Physical Hold Points within the Troubleshooting Plan to limit and control activities.
Southern Nuclear Operating Company Nuclear Management Procedure Problem Solving and Troubleshooting Guidelines NMP-AD-002 Version 10.0 Page 5 of 14
4.0 Responsibilities
4.1 Troubleshooting
Leader- Individual assigned to develop the troubleshooting plan, coordinate work and team discussions, act as a single point of contact and/or obtain changes to the plan, as assigned by the responsible department manager. Perform Just In Time Risk Assessment as described in NMP-DP-001, Operational Risk Awareness.
4.2 Operations
Department The Operations Department, under the direction of the Operations Manager, is responsible for ensuring the troubleshooting activities are supported by:
Approve the Troubleshooting plan where risk has been assessed as Medium or High Providing personnel in support of the Troubleshooting Team Maintain approved system status during troubleshooting activities (i.e. Out Of Service) 4.3 Work Planning The group developing the Troubleshooting plan will determine the level of risk associated with the Troubleshooting plan by using Procedure NMP-DP-001, Operational Risk Awareness.
The plan should consider elimination of worst case, long lead time components early in the process, as potential causes.
Troubleshooting plan steps that alter the configuration of the plant will be implemented and controlled by a planned work order or use of referenced instructions from an approved procedure. This requirement may be waived by the operations shift manager in which case configuration changes will be controlled using detailed instructions in the troubleshooting plan and will be approved by operations before implementation.
Troubleshooting plan steps that do not affect plant configuration control, for example system walkdown, data collection and trending, field observation, and other similar fact finding steps may be implemented by the troubleshooting plan steps.
4.4 Maintenance
Department The Maintenance Department, under the direction of the Maintenance Manager, is responsible for:
Maintenance Manager or his designee will approve High or medium risk Troubleshooting activities where personal safety or economic safety are assessed as Medium or High risk Providing personnel in support of the Troubleshooting Team Ensuring that Troubleshooting Plan is performed and documented in accordance with approved site procedures and safe work practices Determining the need for additional support for troubleshooting activities
- 1. G2.2.25 001/LOIT AND LOCT/SRO/M/F 3.2/4.2/G2.2.25/LO-LP-39209-02//HL-17 AUDIT/
Given the following:
- Unit 1 is at 100% reactor power. - The following RW ST parameters are recorded:
Temperature is 47°F.Level is 93%.
Which one of the following completes the following statement?
Tech Spec action is required for RWST __(1)__,
and the Tech Spec Basis for this param eter limit is to ensure __(2)__.
(1) Level (2) sufficient borated water to support the ECCS during the injection phase of a design basis main steam line break (1) Level (2) sufficient borated water to support the ECCS during the injection phase of a design basis loss of coolant accident (1) Temperature (2) that the amount of cool ing provided from th e RWST during the heatup phase of a main steam line break is consistent with safety analysis assumptions (1) Temperature (2) that the amount of co oling provided from the RW ST during the heatup phase of a main feed line break is consistent with sa fety analysis assumptions A.B.C.D.K/A 2.2.25 Knowledge of the bases in Techni cal Specifications for limitingconditions for operations and safety limits.K/A MATCH ANALYSISThe question asks the candidate straight forward and RO level question of which parameter will place the unit in a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> shutdown LCO, RWST le vel or temperature.
Once determining RWST level is the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> shutdown, the SRO portion requires the candidate to know the correct bases fo r the RWST level.Tuesday, February 25, 2014 11:49:11 AM 1
EXPLANATION OF REQUIRED KNOWLEDGE Per TS SR 3.5.4.1, the RWST borated water temper ature must be maintained greater than or equal to 44° F and less than or equal to 116°F.
Per TS 3.5.4 Bases, the maximum temperature ensures that the amount of cooli ng provided from the RWST during the heatup phase of a f eedline break is consistent with the safety analysis assumptions. The minimum temperature is an assumption in both the MSLB and inadvertent ECCS actuation.
The inadvertent ECCS actuat ion is typically non-limiting.
Per TS SR 3.5.4.1, the RWST borated water must be maintained greater than or equal to 686,000 gallons. Per Te ch Spec rounds O SP 14000-1, 686,000 ga llons correspondsto an indicated level of 94%. Per TS 3.
5.4 Bases, the RWST volume is an explicit assumption for LOCA events. The desired volume limit is set by LOCA and containment analyses.
The volume is not an explicit assumption for non-LOCA events since the required volume is a small fraction of the available volume. The deliverablevolume is different from the total volume c ontained since, due to the design of the tank, more water can be maintained than can be delivered.ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. The first part is correct. 93% level will place the unit in a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> shutdown.The second part is incorrect.
RWST level ensures sufficient injection volume for a DBA LOCA, not a DBA MSLB. However, Safety Injection is required to mitigate a main steam line break and requires suction from the RWST , but is not a limiting factor since the required volume woul d be a small fraction of that available.B. Correct. The first part is correct. See the first part of choice A above.
The second part is correct. Per TS 3.5.4 Bases, the RWST volume is an explicit assumption for LOCA eventsC. Incorrect.Plausible. The first part is incorrect. Per TS SR 3.5.4.1, the RWST borated water temperature must be mainta ined greater than or equal to 44°F and less than or equal to 116°F.
The second part is incorrect.
Per TS 3.5.4 Bases, the maximum temperature ensures that the amount of cooling provided from the RWST during the heatup phase of a feedline break is consistent with the safety analysis assumptions, not a
main steamline break.D. Incorrect.Plausible. The first part is incorrect.
See the first part of choice C above.
The second part is the correct for the first part. Per TS 3.5.4 Bases, the maximum temperatur e ensures that the amount of cooling provided from the RWST during the heatup phase of aTuesday, February 25, 2014 11:49:11 AM 2
feedline break is consistent with the safety analysis assumptions, not a main steamline break.SRO-ONLY JUSTIFICATION(2) Facility operating limitations in the technical specifications and their bases.-Can question be answered solely by knowing = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action? No, the question requires know ledge of the TS Bases for the limit.-Can question be answered solely by knowing the LCO/TRM information listed "above-the-line?"
No, the question is not related to above-the-line Tech Spec. The information is found in a survelliance and TS Bases.
-Can question be answered solely by knowing the TS Safety Limits?
No, thequestion is not related to any Tech Spec Safety Limit.-Does the question involve one or more of the following for TS,TRM, or ODCM?
- Application of Required Actions (Sec tion 3) and Surve illance Requirements (Section 4) in accordance with rules of applicati on requirements (Section 1)* Application of generic LCO requirements (LCO 3.0.1 thru 3.0.7 and SR 4.0.1 thru 4.0.4)* Knowledge of TS bases that is required to analyze TS required actionsand terminology.
Yes, the question asks th e bases for a Tech Spec survelliance.Tuesday, February 25, 2014 11:49:11 AM 3
Level: SROTier # / Group # T3 K/A# G2.2.25 Importance Rating: 3.2 / 4.2 Technical
Reference:
TS 3.5.4, Ammendment No. 96, page 3.5.4-1 & 2 TS Bases 3.5.4, Rev 0, pages B3.5.4-3 & 4
OSP 14000-1, Rev 88.1, pages 8 & 15References provided: None
Learning Objective:
LO-LP-39209-01 For any given it em in section 3.5 of Tech Specs, be able to:
- a. State the LCO.
- b. State any one hour or less required actions.LO-LP-39209-02 Given a set of the Tech Specs and the bases, determine for a specific set of
plant conditions, equipment availability, and operational mode:
- a. Whether any Tech Spec LCOs of section 3.5 are exceeded.
- b. The required actions for all section 3.5 LCOs.LO-LP-39209-03 Describe t he bases for any given Tech Spec in section 3.5.Question origin: BANK -
LOIT Question # 006G2.2.39 001Cognitive Level: M/F 10 CFR Part 55 Content: 41.8 / 43.2 Comments:
You have completed the test!Tuesday, February 25, 2014 11:49:11 AM 4
RWST 3.5.4 Vogtle Units 1 and 2 3.5.4-1 Amendment No. 96 (Unit 1)
Amendment No. 74 (Unit 2)
3.5 EMERGENCY
CORE COOLING SYSTEMS (ECCS)
3.5.4 Refueling
Water Storage Tank (RWST)
LCO 3.5.4 The RWST shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RWST boron concentration not within
temperature not within
limits. A.1 Restore RWST to OPERABLE status.
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> B. One or more sludge mixing pump isolation
valves inoperable.
B.1 Restore the valve(s) to OPERABLE status.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> C. Required Action and associated Completion
Time of Condition B not
met. C.1 Isolate the sludge mixing system. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> D. RWST inoperable for reasons other than
Condition A or B.
D.1 Restore RWST to OPERABLE status.
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (continued)
RWST 3.5.4 Vogtle Units 1 and 2 3.5.4-2 Amendment No. 158 (Unit 1)
Amendment No. 140 (Unit 2)
ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME E. Required Action and associated Completion
Time of Condition A or D
not met. E.1 Be in MODE 3.
AND E.2 Be in MODE 5.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.4.1 --------------------------NOTE-------------------------------
Only required to be performed when ambient air temperature is < 40
°F. -----------------------------------------------------------------
Verify RWST borated water temperature is 44°F and 116°F.
In accordance with the Surveillance Frequency Control Program SR 3.5.4.2 Verify RWST borated water volume is 686,000 gallons. In accordance with the Surveillance Frequency Control Program SR 3.5.4.3 Verify RWST boron concentration is 2400 ppm and 2600 ppm.
In accordance with the Surveillance Frequency Control Program SR 3.5.4.4 Verify each sludge mixing pump isolation valve automatically closes on an actual or simulated
RWST Low-Level signal.
In accordance with the Surveillance Frequency Control Program RWST B 3.5.4 Vogtle Units 1 and 2 B 3.5.4-3 Revision No. 0 BASESBACKGROUND reduction of SDM or excessive boric acid precipitation in the core (continued) following the LOCA, as well as excessive stress corrosion of mechanical components and systems inside the containment. APPLICABLE During accident conditions, the RWST provides a source of SAFETY ANALYSES borated water to the ECCS and Containment Spray System pumps. As such, it provides containment cooling and depressurization, core cooling, and replacement inventory and is a source of negative reactivity for reactor shutdown (Ref. 1). The design basis transients and applicable safety analyses concerning each of these systems are discussed in the Applicable Safety Analyses section of B 3.5.2, "ECCS-Operating"; B 3.5.3, "ECCS-Shutdown"; and B 3.6.6, "Containment Spray and Cooling Systems." These analyses are used to assess changes to the RWST in order to evaluate their effects in relation to the acceptance limits in the analyses. The RWST must also meet volume, boron concentration, and temperature requirements for non-LOCA events. The volume is not an explicit assumption in non-LOCA events since the required volume is a small fraction of the available volume. The deliverable volume limit is set by the LOCA and containment analyses. For the RWST, the deliverable volume is different from the total volume contained since, due to the design of the tank, more water can be contained than can be delivered. The minimum boron concentration is an explicit assumption in the main steam line break (MSLB) analysis to ensure the required shutdown capability. The maximum boron concentration is an explicit assumption in the inadvertent ECCS actuation analysis, although it is typically a nonlimiting event and the results are very insensitive to boron concentrations. The maximum temperature ensures that the amount of cooling provided from the RWST during the heatup phase of a feedline break is consistent with safety analysis assumptions; the minimum is an assumption in both the MSLB and inadvertent ECCS actuation analyses, although the inadvertent ECCS actuation event is typically nonlimiting. The MSLB analysis has considered a delay associated with the interlock between the VCT and RWST isolation valves, and the (continued)
RWST B 3.5.4 Vogtle Units 1 and 2 B 3.5.4-4 Rev. 1-10/01 BASESAPPLICABLE results show that the departure from nucleate boiling design SAFETY ANALYSES basis is met. The delay has been established as 27 seconds, (continued) with offsite power available, or 39 seconds without offsite power (includes 12 seconds for the Emergency Diesel Generator). This response time includes an electronics delay, a stroke time for the RWST valves, and a stroke time for the VCT valves. For a large break LOCA analysis, the minimum water volume limit of 499,091 gallons and the lower boron concentration limit of 2400 ppm are used to compute the post LOCA sump boron concentration necessary to assure subcriticality. The large break LOCA is the limiting case since the safety analysis assumes that all control rods are out of the core. The upper limit on boron concentration of 2600 ppm is used to determine the maximum allowable time to switch to hot leg recirculation following a LOCA. The purpose of switching from cold leg to hot leg injection is to avoid boron precipitation in the core following the accident. In the ECCS analysis, the containment spray temperature is assumed to be equal to the RWST lower temperature limit of 44F. If the lower temperature limit is violated, the containment spray further reduces containment pressure, which decreases the rate at which steam can be vented out the break and increases peak clad temperature. (The reduction in containment pressure correspondingly reduces the density of the vented steam. This reduces the flow of steam out of the core, which translates into a decrease in the ECCS flooding rate. This decrease in the flooding rate causes the increase in peak clad temperature.) The upper temperature limit of 116F is used in the small break LOCA analysis and containment OPERABILITY analysis. Exceeding this temperature will result in a higher peak clad temperature, because there is less heat transfer from the core to the injected water for the small break LOCA and higher containment pressures due to reduced containment spray cooling capacity. For the containment response following an MSLB, the lower limit on boron concentration and the upper limit on RWST water temperature are used to maximize the total energy release to containment. The RWST satisfies Criterion 3 of 10 CFR 50.36 (c)(2)(ii). (continued)
Approved By J.B. Stanley Vogtle Electric Generating Plant Procedure Version 14000-1 88.1 Effective Date 06/21/2013 OPERATIONS SHIFT AND DAILY SURVEILLANCE LOGS Page Number 8 of 36 Printed January 21, 2014 at 15:43 Sheet 2 of 10 DATA SHEET 1 MODE 1 & 2 MODE _______________
DATE _______________
LCO TECH SPEC INDICATION LIMIT(S)
METHOD OF VERIFICATION SURV REQ PARAMETER INSTRUMENT DAY NIGHT TOLERANCE LCO/PROC EACH ACCUMULATOR SHALL BE OPERABLE
SR 3.5.1.1 1HS-8808A MLB001 2.3 VERIFY DISCHARGE VALVE POSITION VALVE POSITION (INIT) 1HS-8808B MLB002 2.3 OPEN 3.5.1 1HS-8808C MLB001 2.4 1HS-8808D MLB002 2.4 TWO ECCS FLOW TRAINS SHALL BE OPERABLE
SR 3.5.2.1 1HS-8806 OPEN AND POWER REMOVED VERIFY VALVES POSITIONED AND POWER 1HS-8835 OPEN AND POWER REMOVED REMOVED BY ASSOCIATED LOCKOUT SWITCH LIGHT 1HS-8813 OPEN AND POWER REMOVED EXTINGUISHED AND SWITCH IN LOCKOUT VALVE STATUS (INIT) 1HS-8802A CLOSED AND POWER REMOVED 3.5.2 POSITION 1HS-8802B CLOSED AND POWER REMOVED 1HS-8840 CLOSED AND POWER REMOVED 1HS-8809A OPEN AND POWER REMOVED 1HS-8809B OPEN AND POWER REMOVED ESFAS INSTRUMENTATION SHALL BE OPERABLE SR 3.3.2.1
FCN 7B 1LI-0991A CHANNEL CHECK CHANNEL CHECK RWST LEVEL (%) 1LI-0993A REQUIRED 4 3.3.2(K) ACCIDENT MONITORING INSTRUMENTATION SHALL SR 3.3.3.1 FCN 9 1LI-0990A REQUIRED 2 3.3.3 (B,G,H,J) BE OPERABLE CHANNEL CHECK SR 3.5.4.2 1LI-0992A
>94%
3.5.4 COMPLETED
BY: DAY: TIME: NIGHT: TIME: SS REVIEW: DAY: TIME: NIGHT: TIME:
Approved By J.B. Stanley Vogtle Electric Generating Plant Procedure Version 14000-1 88.1 Effective Date 06/21/2013 OPERATIONS SHIFT AND DAILY SURVEILLANCE LOGS Page Number 15 of 36 Printed January 21, 2014 at 15:43 Sheet 9 of 10 DATA SHEET 1 MODE 1 & 2 MODE _______________
DATE _______________
LCO TECH SPEC INDICATION LIMIT(S)
METHOD OF VERIFICATION SURV REQ PARAMETER INSTRUMENT DAY NIGHT TOLERANCE LCO/PROC CREFS ACTUATION OPERABLE SR 3.3.7.1
FCN 3 CR INTAKE RADIATION 1RE-12116 CHANNEL CHECK
3.3.7 CHANNEL
CHECK MONITORS (INIT) 1RE-12117 REQUIRED 2 FHB ACTUATION OPERABLE TRS 13.3.6.1 FHB EFFL RADIOGAS ARE-2532A
- 13.3.6 CHANNEL CHECK FHB ISO (INIT) ARE-2532B REQUIRED 1 FHB ACTUATION OPERABLE TRS 13.3.6.1 FHB EFFL RADIOGAS ARE-2533A
- 13.3.6 CHANNEL CHECK FHB ISO (INIT) ARE-2533B REQUIRED 1 *INDICATING NORMALLY. ALL STATUS AND ALARM LIGHTS EXTINGUISHED. DG1A FUEL OIL INVENTORY VERIFY FUEL OIL STORAGE TANK LEVEL SR 3.8.3.1 DG 1A LEVEL (%) 1-LI-9024 82% 3.8.3 DG1B FUEL OIL INVENTORY VERIFY FUEL OIL STORAGE TANK LEVEL SR 3.8.3.1 DG 1B LEVEL (%) 1-LI-9025 82% 3.8.3 TWO INDEPENDENT CONTROL ROOM EMERGENCY FILTRATION SYSTEMS SHALL BE OPERABLE VERIFY CONTROL ROOM TEMP SR 3.7.10.1 SR 3.7.11.1
NOTE: TEMPERATURE INDICATION IS OBTAINED FROM HAND-HELD TEST EQUIPMENT. RECORD INSTRUMENT INFORMATION BELOW.
INSTRUMENT ID NO.
N/A CAL DUE DATE CONTROL ROOM TEMPERATURE (F) M&TE <85 F 3.7.10 3.7.11 THE RWST SHALL BE OPERABLE VERIFY TEMPERATURE SR 3.5.4.1 TRS 13.1.7.1 RWST TEMPERATURE (F) 1TIS-10980
>51 F * <109 F
- 3.5.4 13.1.7 *WITH INDICATED RWST TEMPERATURE OUTSIDE THE LIMITS, THEN VERIFY RWST TEMPERATURE IS WITHIN TECHNICAL SPECIFICATION LIMITS BY PLACING THE RWST ON RECIRC USING SLUDGE MIXING PUMP WITH HEATER OFF AND OBSERVING 1-TI-10982 TO BE WITHIN 44F AND 116F. THE ULTIMATE HEAT SINK SHALL BE OPERABLE COMPUTER POINT T2601* <90 F 3.7.9 VERIFY WATER -OR- TEMPERATURE AND LEVEL SR 3.7.9.2 TEMPERATURE (F) 1TJI-1692 POINT 2* COMPUTER POINT T2602* -OR- 1TJI-1692 POINT 17* *IF COMPUTER POINT AND RECORDER POINT ARE NOT AVAILABLE, TEMPERATURE READING MUST BE OBTAINED LOCALLY USING HAND-HELD TEST EQUIPMENT. RECORD INSTRUMENT INFORMATION BELOW.
INSTRUMENT ID NO.
N/A CAL DUE DATE SR 3.7.9.1 LEVEL 1LI-1606 >73% (%) 1LI-1607 CONTAINMENT AIR TEMPERATURE SHALL NOT SR 3.6.5.1 COMPUTER POINT T2501 EXCEED 120F VERIFY AVERAGE AIR TEMPERATURE (F) COMPUTER POINT T2502 NA TEMPERATURE COMPUTER POINT T2503 COMPUTER POINT UT2501 (AVG)
<120 F 3.6.5 *IF COMPUTER POINT IS NOT AVAILABLE VERIFY CNMT HI TEMP ALARM ALB-01 (E06) IS NOT IN ALARM. ALB-01 (E06) NOT IN ALARM
- IF COMPUTER POINT AND ALB-01 (E06) ARE NOT AVAILABLE, TEMPERATURE READING MUST BE OBTAINED LOCALLY USING HAND-HELD TEST EQUIPMENT FOR 1TE-2612 FOR POINT T2502 AND 1TE-2613. FOR POINT T2503 RECORD INSTRUMENT INFORMATION BELOW. USE MCB INDICATOR 1TI-2563 FOR POINT T2501 AND AVERAGE THE THREE.
INSTRUMENT ID NO.
<120 F CAL DUE DATE COMPLETED BY: DAY: TIME: NIGHT: TIME: SS REVIEW: DAY: TIME: NIGHT: TIME:
- 1. G2.3.4 001/LOCT AND LOIT/SRO/M/F 3.2/3.7/G2.3.4/LO-LP-40101-08///
Initial condition:
- General Em ergency has been declared.
Current conditions:
- A first responder is br iefed to rescue an injured worker.
- Health Physics estimate s the first responder will receive 11 rem TEDE dose while performing the rescue.
Which one of the following completes the following statement?
Per 91301-C, "Emergency Exposure Guideli nes," the dose rece ived by the first responder during the rescue __(1)__ be added to the responder's occupational non-emergency exposure, and the LOWEST level of approv al required to authorize the first responder's rescue exposure is the __(2)__. __(1)__ __(2)__ will Health Physics Supervisor will Emergency Director will NOT Health Physics Supervisor will NOT Emergency Director A.B.C.D.K/A 2.3.4 Knowledge of radiation exposure limits under normal or emergencyconditions.K/A MATCH ANALYSIS The question sets up a plausible scenario which includes all the required KA elements.
First the SRO candidate must determine if the exposure in the General Emergency would be added to normal expo sure already accumulated and the authorization level required for the given dose.EXPLANATION OF REQUIRED KNOWLEDGE Per Admin procedure 91301-C, "Emergency Exposure Guidelines" NOTE on TABLE 1, dose to workers performing emergency servic es may be treated as a once-in-a-lifetimeTuesday, February 25, 2014 1:44:48 PM 1
exposure and should not be added to occu pational exposure accumulated under non-emergency conditions. Per Responsibilities 2.1, the Emergency Director (ED) hasthe sole authority to allow radiation exposures in excess of 10CFR20 limits. Per Responsibilities 2.2.4, the HP Superviso r or designedd can authorize individuals to recieve radiation exposures in excess of VEGP Admin Gui delines, but not in excess of 10CFR20 limits.ANSWER / DISTRACTOR ANALYSIS A. Incorrect. Plausible. Pa rt 1 is incorrect however 'pl ausible' since it's reasonable to assume the candidate may dete rmine that all radiation exposure for the year would be additive to ensure accurate accounting for health reasons.
Part 2 is incorrect however 'pl ausible' since the candidate may determine, based on plant conditi ons, that with the exposureless than 25 Rem, that Emerg ency Director involvement would not be required and th erefore the Health Physics Supervisor could authorizes this.B. Incorrect. Plausible. Part 1 is incorrect. See Part 1 of choice A above.
Part 2 is correct. Per 91301-C 'EMERGE NCY EXPOSUREGUIDELINES', the Emergency Director (ED) has the sole
authority to allow radiation exposures in excess of 10CFR20 limits.C. Incorrect. Plausible. Part 1 is correct. Per 91301-C 'EMERGENCY EXPOSUREGUIDELINES', dose to worker s performing emergency services may be treated as an once-in-a-lifetime exposure and should not be added to occupational exposure accumulated under non-emergency conditions.
Part 2 is incorrect. See Part 2 of choice A above.D. Correct. Part 1 is correct.
See Part 1 of choice C above..Part 2 is correct. See part 2 of choice B abvoe.SRO JUSTIFICATION (10CFR43(b))(5) Assessment of facility conditions and selection of appropriate proceduresduring normal, abnormal, and emergency situations.-Can the question be answered solely by knowing "systems knowledge", i.e.,how the system works, flowpath, logic, component location? No, the knowledgerequired pertains to administrative guidance and ED n on deligable duties.-Can the question be answered solely by knowing immediate operator actions? No, information found in IOAs is not involved in the question.
-Can the question be answered solely by knowing entry conditions for AOPs orplant parameters that require direct entry to major EOPs?
No, EOPs and AOPsTuesday, February 25, 2014 1:44:48 PM 2
are not involved with this questioon.-Can the question be answered solely by knowing the purpose, overall sequenceof events, or overall mitigati ve strategy of a procedure? No, detailed knowledgeof process and responsibilities within an admin procedure are required. Overallknowledge will not answer the question.-Does the question re quire one or more of the following?* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or se ction of a procedure to mi tigate, recover, or with which to proceed* Knowledge of when to impl ement attachments and appendices, including how to coordinate these it ems with procedure steps* Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific s ubprocedures or emergency contingency procedures* Knowledge of administrative procedures that specify hierarchy,implementation, and/or coordination of plant normal, abnormal, and emergency procedures. Yes, specific knowledge of the duties of the EDand HP Supervisor as well as how emergency exposures are recorded fordose purposes is required.
Level: SROTier # / Group # T3 K/A# G2.3.4 Importance Rating: 3.2 / 3.7Technical
Reference:
ADMIN 9130 1-C, Rev 12.1, pages 3 & 9References provided: None Learning Objective: LO-LP-40101
-35 State what group of people should be first considered for emer gency exposure, and what group should not be allowed to
receive an emergency exposure
(91301-C). (SRO only)
LO-LP-40101-08 State from memo ry ED duties that cannot be delegated (SRO only).Question origin: MODIFI ED - Turkey Point 2011 NRC Question # G.2.3.4Cognitive Level: M/F 10 CFR Part 55 Content: 43.5 Comments: You have completed the test!Tuesday, February 25, 2014 1:44:48 PM 3
Approved By S. C. Swanson Vogtle Electric Generating Plant Procedure No. Version 91301-C 12.1 Effective Date 03/12/2013 EMERGENCY EXPOSURE GUIDELINES Page Number 3 of 16 Printed October 4, 2013 at 12:52 INFORMATION USE
1.0 PURPOSE
The purpose of this procedure is to provide instructions and controls for radiation exposures in excess of the Vogtle Electric Generating Plant (VEGP) Administrative Guidelines, or in excess of the 10CFR20 occupational limits during emergency conditions.
2.0 RESPONSIBILITIES
2.1 The Emergency Director (ED) has the sole authority to allow radiation exposures in excess of 10CFR20 limits in accordance wit h the provisions of this procedure.
2.2 The Health Physics (HP) Supervisor, or designee, shall have the following responsibilities:
2.2.1 Preparing
Permits for Emer gency Radiation Exposure (PERE).
(1985304698)
2.2.2 Maintaining
records of emer gency exposures for each individual.
2.2.3 Providing
recommendations to the ED on exposure control measures including issuance of Dosimetry, use of protective equipment and issuance of thyroid blocking agents such as potassium iodide (KI).
2.2.4 Authorizing
individuals to receive radiation exposures in excess of the VEGP Administrative Guidelines, but which do not exceed the 10CFR20 limits.
3.0 PREREQUISITES
An emergency situation exists which results in a need to initiate corrective actions, protective ac tions, sampling activities, or lifesaving measures which might result in exposures greater than 10CFR20 limits. 4.0 PRECAUTIONS
4.1 Personnel
authorized to receive exposures in excess of 10CFR20 limits shall meet t he following criteria:
(1985305022) 4.1.1 Personnel shall be familiar with the risks of exposure to the higher radiation levels which are likely during emergency conditions as outlined in Table 2 and Table 3.
Approved By S. C. Swanson Vogtle Electric Generating Plant Procedure No. Version 91301-C 12.1 Effective Date 03/12/2013 EMERGENCY EXPOSURE GUIDELINES Page Number 9 of 16 Printed November 19, 2013 at 15:22 TABLE 1 EMERGENCY EXPOSURE GUIDELINES (1985305256) (1985305827)
NOTES Dose limits listed in this table apply to doses incurred over the duration of the emergency. Dose to workers performing emergency services may be treated as an once-in-a-lifetime exposure and should not be added to occupational exposure accumulated under non-emergency conditions. Workers performing services during emergencies shall limit dose to the lens of the eyes to three times the listed value and doses to any other organ (including skin and body extremities) to ten times the listed value.
Dose Limit (REM)
Total Effective Dose Equivalent Activity Condition 5 All 10 Protecting Valuable Property Lower Dose not practicable 25 Lifesaving or protection of large population Lower Dose not practicable
>25 Lifesaving or protection of large population Only on a voluntary basis to persons fully aware of the
risks involved
- 1. G2.3.7 001/LOIT AND LOCT/SRO/M/F 3.5/3.6/G2.3.7/LO-LP-63920-03///
Given the following:
- A Systems Operator (SO) will make multiple entries into AB-A-33 to place the CVCS cation demineralizer in service.
- The SO will use RWP 14-0108 (red RWP).
- The SO will exceed the A nnual Administrative 4000 mrem per year TEDE limit during the task.
Which one of the following completes the following statement?
The SO __(1)__ required to re ceive an ALARA briefing pr ior to each AB-A-33 entry, and per NMP-HP-001, "Radiation Pr otection Standard Practices," the __(2)__ is the LOWEST level of approval required to exceed the Ad ministrative dose limit. __(1)__ __(2)__ is HP Manager is Plant Manager is NOT HP Manager is NOT Plant Manager A.B.C.D.Tuesday, February 25, 2014 2:59:06 PM 1
K/AG2.3.7 Radiation ControlAbility to comply with radiation work permit requirements duringnormal or abnorm al conditions:K/A MATCH ANALYSIS:
The candidate is presented with a scenario where an Auxiliary Building Operator is required to enter several rooms with high dos e rates in the area.
The Operator is also on the verge of exceeding his Annual TEDE limits of 40 00 mrem. The candidate has to determine the minimum level of authority that may approv e exceeding the annual TEDE limits.EXPLANATION OF REQUIRED KNOWLEDGE Per NMP-HP206 step 5.3.31.
2 states that RED RWPs are "Single Use" type briefings.
Step 5.3.11.1.4 defines Sing le Use as "individuals must be authorized and the authorization is good for one entry only." Per NMP-HP-001 step 6.2.3, t here are 3 different administrat ive levels that require 3 different levels of approval:
- 1. 2000 mrem in a year requires HP Supervisor, Physi cist, or Manager approval.2. 4000 mrem in a year requres AGM or Plant General Manager approval.3. 4500 mrem in a year requires Project Vice President approval.
Dose in excess of 5mrem requires special NRC approva l for normal operation, ED approval during emergencies. These di fferent admin limit s do not have a noun description/name and therefore are generally referred to by the associated dose limit.
DISTRACTOR ANALYSIS: A. Incorrect. Plausible.
Part 1 is correct. RWP 14-0108 is a red RWP and therefore requires a Single Use briefing on each entry per NMP-HP-206.
Part 2 is incorrect. Per NM P-HP-001, exceedi ng the 4000mrem per year admin dose limit will required Plant General Manager approval. However, the HP Supervisor can approval all dose up to the 4000mrem limit.
Therefore, this dist ractor is plausible.B. Correct. Part 1 is correct.RED RW P and requires a briefing for each entry. The minimum authority level to approve the TEDE extension is the Pl ant Manager from t he choices presented.
Part 2 is correct. Per NMP-HP-001, exceeding the 4000mrem per year admin dose limit will required Plant General Manager approval.Tuesday, February 25, 2014 2:59:48 PM 1
C. Incorrect. Plausible. Part 1 is incorrect. RWP 14-0108 is a red RWP and therefore requires a Single Use briefing on each entry per NMP-HP-206.
However, both yellow and green RW Ps allow re-entry into theroom with a single HP brief.
Therefore, this distractor is plausible.
Part 2 is incorrect. See Part 2 of choice A above.
D. Incorrect. Plausible. Part 1 is in correct. See Part 1 of choice C above.
Part 2 is correct. See Part 2 of choice B above.SRO JUSTIFICATION(4) Radiation hazards that may arise during normal and abnormal situations,including maintenance activities and various contamination conditions. Yes, specific knowledge of administrative procedures associated withradiological safety and rad exposure authorization levels during normal plantconditions is tested.
Level: SROTier # / Group # T3 K/A# G2.3.7 Importance Rating: 3.5 / 3.6 Technical
Reference:
NMP-HP-001, Rev 5.2, page 14 & 15 NMP-HP-206, Rev 3.0, pages 8 &12
V-LO-LP-63930, page 12References provided: None
Learning Objective:
LO-LP-63930-06 Stat e the entry requirements applicable to each of the following:
- b. Radiation Control Area (RCA)
- c. Radiation Area
- e. Locked High Radiation Area LO-LP-63920-03 State th e plant administrative limits/guidelines for radiation dose.
LO-LP-63920-04 State the actions to be taken if administrative dose limits are being
approached.Question origin: MODIFI ED - HL18 NRC - G2.3.7Cognitive Level: M/F
10 CFR Part 55 Content: 41.12 / 43.4You have completed the test!Tuesday, February 25, 2014 2:59:48 PM 2
- 1. G2.3.7 003/LOCT AND LOIT/SRO/M/F 3.5/3.6/G2.3.7///HL18 NRC/079G2.1.27(Original Question from HL18 NRC)
Given the following:
- A Fuel Handling Coordinator (F HC) is entering the Spent Fuel Pool area.
- The FHC is reviewing his RWP prior to beginning work and notices an ALARA briefing is required.
- The dose rate is 900 mrem/hour due to damaged fuel assemblies.
- The FHC will also exceed 2000 mrem Annual TEDE limits while in the area.
Which one of the following completes the following statement?
Based on the area dose rate, the FHC will be required to receive an ALARA briefing prior to __(1)__ entry, and per NMP-HP-001, "Radiation Protection Standard Practice s", the ___(2)___ is the MINIMUM authority level required to exceed the Annual TEDE limit.
__(1)__ __(2)__ each HP Manager each Plant General Manager ONLY the first HP Manager ONLY the first Plant General Manager A.B.C.D.Friday, November 15, 2013 3:12:19 PM 1
Southern Nuclear Operating Company Nuclear Management Procedure Radiation Protection Standard Practices NMP-HP-001 Version 5.2 Page 14 of 16 6.1.55 20.2202 Notification of Incidents Paragraph (a)(2) and (b)(2), for reporting the release of radioactive material inside or outside of a restricted area is caveated as not applicable to, "locations where personnel are not normally stationed during routine operations, such as hot cells or process enclosures." For purposes of this section, consistent with the answer to question 56 of the first NRC 10CFR20 Q & A document, locations where personnel are not normally stationed will be interpreted as areas, rooms and enclosures which are not normally occupied nor periodically patrolled during normal plant operations and maintenance.
6.1.56 20.2203 Reports of Exposures, Radiation Levels and Concentrations of Radioactive Material Exceeding the Limits No fleet practices identified.
6.1.57 20.2204 Reports of Planned Special Exposures No fleet practices identified.
6.1.58 20.2206 Reports of Individual Monitoring The intent of Regulatory Guide 8.7, "Instructions for Recording and Reporting Occupational Radiation Exposure Data," will be met in complying with this paragraph with NRC Form 5.
6.1.59 20.2301 Applications for Exemptions No fleet practices identified.
6.1.60 20.2302 Additional Requirements No fleet practices identified.
6.1.61 20.2401 Violations No fleet practices identified.
6.2 Other
Consensus Positions
6.2.1 Whole
Body Count Performance Frequency Monitored workers will be given an entrance and exit whole body count (WBC) or whole body scan (WBS). The exit WBC from another SNC site can be used in lieu of an entrance WBC if the SNC site was the last site the worker entered an RCA and/or
monitored. Upon request from a worker, WBCs will be provided to the worker on a voluntary and reasonable basis.
6.2.2 Dose Limits for Workers Who Provide Outage Support at a SNC Plant Other Than Their Home Plant.
Workers should be limited to 500 mrem per visit, unless express consent is given by the home plants management to exceed that limit.
6.2.3 Administrative
Annual TEDE Dose Limits and the Approval Authority Necessary to Exceed Limits 6.2.3.1 2000 mrem in a year requires HP Support Supervisor, Plant Health Physicist, or HP Manager approval.
Southern Nuclear Operating Company Nuclear Management Procedure Radiation Protection Standard Practices NMP-HP-001 Version 5.2 Page 15 of 16 6.2.3.2 4000 mrem in a year requires AGM or Plant General Manager approval.
6.2.3.3 4500 mrem in a year requires Project Vice President approval.
6.2.4 Discrepant
Dosimeter Investigation Criteria An assessment of worker's dose should be initiated for discrepant dosimetry results when the following criteria are met: the primary or secondary dosimeter dose exceeds 100 mrem; and the secondary dosimeter reading differs by more than 25% from the primary dosimeter.
6.2.5 Dose Monitoring Threshold All individuals entering an RCA will be monitored for radiation exposure. A single dosimeter suffices for Visitors or Radiation Workers whose annual dose from sources external to the body is not expected to exceed 100 mrem at a particular station. An Optically Stimulated Luminescent Dosimeter (OSLD) and a self-reading dosimeter, such as an Electronic Dosimeter (ED), will be provided to all other individuals.
6.2.6 Training
Requirements for Visitors or Temporary Radiation Workers Who Enter RCAs All SNC plants will administer training to all who must enter the RCAs in the following manner: 6.2.6.1 If the individual is expected to receive < 100 mrem in a year, the individual will be escorted by a GET qualified worker and will be provided instructions.
6.2.6.2 If the individual is expected to receive >100 and <500 mrem in a year, the individual will receive a visitor handout (containing all of the instruction elements required by 10CFR19.12) and will acknowledge receipt of the instructions or handout by signing a form. The individual will have a GET-trained radiological escort. If special circumstances dictate that entries into contaminated areas are required, the individual will receive dress-out training (if he has no history of such training at our plants). Similarly, if the individual requires entry to high radiation areas, special training or instructions may be required.
6.2.6.3 If the individual is expected to receive >=500 mrem in a year, the individual is required to complete GET which includes testing. If the individual has previous GET training at a nuclear facility within the past two years or as allowed by the Training Department, then exemption GET and the exemption GET test can be administered.
6.2.7 Air Flow for Fume Hoods Containing Radioactive Materials/Fluids Sample station fume hoods containing radioactive materials/fluids will meet a minimum air flow acceptance criteria of 100 LFPM, and will be labeled to ensure that 100 LFPM is maintained or exceeded.
6.2.8 Respirator
Training and Fit Test Frequencies Sites will apply a program whereby classroom or Computer-based training with an examination will be conducted annually. For individuals required to utilize respirator Southern Nuclear Operating Company Nuclear Management Procedure Issuance, Use and Control of Radiation Work Permits NMP-HP-206 Version 3.0 Page 8 of 32 5.3.11 Enter the End Date in the same manner as step 5.3.10. When the End Date is entered, press the "Tab" key to advance to the "Authorization Type" field.
5.3.11.1 The type of worker authorization is based upon the type of RWP being written. Worker authorizations fall into 4 categories:
5.3.11.1.1 All - Anyone can use this RWP.
5.3.11.1.2 Work Group - Work Group must be authorized.
5.3.11.1.3 Individual - Individuals must be authorized.
5.3.11.1.4 Single Use - Individuals must be authorized and the authorization is good for one entry only.
5.3.12 From the pull-down menu, select "All" or "Work Group" for a Green RWP. Select "Individual" for a Yellow RWP. Select "Single Use" for a Red RWP.
5.3.13 Press the "Tab" key to advance to the "RWP Type" field.
5.3.14 With the cursor in the "Type" field, select the appropriate RWP type (General or Specific) from the pull-down table. Press the "Tab" key to advance to the "Principle
Work Document" field.
5.3.15 With the cursor in the "Principle Work Document" field, type the activity or MWO number, if applicable. If the RWP is not written for a specific activity or MWO, this field may be either left blank or type N/A in the field.
5.3.16 Press the "Tab" key and advance past the ALARA Review Number field. Tab to the HP Job Coverage field.
5.3.17 In the HP Coverage field, enter None, Intermittent, or Continuous as appropriate for the RWP. Table 1 defines the type of HP coverage used.
5.3.18 Press the "Tab" key to advance to the "Job Description" field.
5.3.19 Type a short general description of the work to be performed in the "Job Description" field. Press "Tab" to advance to the "Location" field.
NOTE For an RWP, Location is the area where the majority of the work should be performed.
5.3.20 Type the work location code or select a location from the pull-down table in the "Location" field.
5.3.21 When the work location is selected, press "Tab" to advance to the "Area" field.
5.3.22 The Area field is normally left blank or N/A is typed in this field. Press "Tab" to advance to the "Comments" field. Enter comments as needed or leave blank.
Southern Nuclear Operating Company Nuclear Management Procedure Issuance, Use and Control of Radiation Work Permits NMP-HP-206 Version 3.0 Page 12 of 32 5.3.29.2 When "Detail" is selected, a Dosimetry Types dialog box will appear. Check the box next to the type of dosimetry required for the RWP. Click on the "OK" button.
5.3.29.3 Click on the "Apply" button at the bottom of the screen.
5.3.30 Select the "Worker Instructions" tab in the "Maintain RWP" screen.
5.3.30.1 With the cursor in the blank Worker Instructions field, click on the right mouse button and select "Detail" from the pop-up table.
5.3.30.2 When Detail is selected, a "Worker Instructions" dialogue box will appear. Place a check mark next to the applicable work instructions, and select "OK".
5.3.30.3 If working from a model, "Rem_occ" may be used to remove any instructions that may be no longer necessary. To remove an instruction, place the cursor in the line of the instructions and click on the "Rem_occ" button.
5.3.30.4 Worker Instructions may be added as free form text.
5.3.30.5 Additional Worker Instructions may be added by using the "Add_occ" button.
5.3.30.6 When all Worker Instructions are entered, click on "Apply" at the bottom of the screen.
5.3.31 Select the "Briefing" tab in the "Maintain RWP" screen.
5.3.31.1 With the cursor in the "Briefing" field, right click on the mouse and select "Detail" from the pop-up table.
5.3.31.2 When Detail is selected, a dialogue box will open. Check the appropriate type of briefing for the RWP and select "OK".
For Red RWPs, use "Single Use" briefing type. For Yellow RWPs, the briefing type is conditional based on the activity. For Green RWPs, no briefing is required.
5.3.31.3 Select "Yes" or "No" in the "Required" field if applicable.
5.3.31.4 After entering the briefing type, click on the "Apply" button at the bottom of the screen.
5.3.32 Select the Supervisors tab in the Maintain RWP screen.
5.3.32.1 Type the name of the job supervisor in the "Job Supervisor" field. Press the "Tab" key.
5.3.32.2 In the "Department" field, select the appropriate department from the pull-down table. Press the "Tab" key.
5.3.32.3 In the "Phone/Ext" field, type the phone number of the job supervisor. Press the "Tab" key.
V-LO-LP-63930 III. LESSON OUTLINE NOTES 12 of 22 Risk-Based RWP Format and Requirements Color Code Radiological Significance Types of RWPs Type of Briefing Required General Radiological Conditions Green Low All General RWPs And Specific RWPs with low radiological risk No ALARA briefing required. Dose Rate: < 100 mrem/hr Contamination Levels: < 200,000 dpm/100 2Airborne Levels: < 0.3 DAC
- Workers should always refer to the most recent survey information for the area(s) being worked in.
Yellow Moderate Specific RWPs that are tied to unique Work Groups - Specific RWPs covering work in areas with intermediate levels of radiological risk. Initial ALARA briefing required prior to first entry. Additional ALARA briefing required when specified rad conditions are exceeded.
A pre-job ALARA briefing will be required if: Radiological conditions that are addressed in the Worker Instructions section may be exceeded, or If the RWP default settings for the accumulated dose or dose rate alarms may be exceeded, or Breach of a contaminated system Dose Rate: < 1000 mrem/hr Contamination Levels: < 500,000 dpm/100 cm 2 Airborne Levels: < 0.3 DAC
- Workers should always refer to the most recent survey information for the area(s) being worked in.
Red High Specific RWPs covering work in areas with high levels of radiological risk. ALARA Briefing required prior to each entry. Radiological conditions on the RWP will be based on actual, projected or historical survey information. Latest rad conditions and specific instructions will be covered in the pre-job ALARA briefing Dose Rate: > 1000 mrem/hr Contamination Levels: > 500,000 dpm/100 cm 2 Airborne Levels: > 0.3 DAC
- Workers should always refer to the most recent survey information for the area(s) being worked in.
- 6. Normally, HP will survey the work area prior to issuing an RWP a) In high radiation areas , survey performance may not be consistent with ALARA
- 1. G2.4.46 001/LOIT AND LOCT/SRO/C/A 4.2/4.2/G2.4.46/LO-TA-40002///056AG2.4.45At time 1000:
- Unit 1 is in Mode 6.At time 1005 the following alarms illuminate:
- ALB32-D02 RESV AUX XF MR 1NXRA HI SIDE PHOC LOR TRIP
- ALB32-E02 RESV AUX XF MR 1NXRB HI SIDE PHOC LOR TRIP
- ALB35-A10 DG1A TRIP OVERSPEED
- ALB35-F10 DG1A EMERGENCY START
- ALB36-A01 4160V SWGR 1AA02 TROUBLE
- ALB37-A01 4160V SWGR 1BA03 TROUBLE alarms , then subsequently clears.
- ALB38-F10 DG1B EMERGENCY STARTCurrent time is 1025:
Based on the current time, which one of the following is the correct Emergency Classification required to be declared?REFERENCE PROVIDED Alert Emergency (CA3)
Alert Emergency (SA5)
Notification of U nusual Event (SU1)
Notification of U nusual Event (CU3)
A.B.C.D.G2.4.46 Emergency Procedures / PlanAbility to verify that the alarms are consistent with the plantconditions.K/A MATCH ANALYSIS:
The candidate must analyze various alarms and indications a ssociated with theelectrical distribution system to determine plant status and the correct emergency classification. The event initiation ti me will also affect the classification.EXPLANATION OF REQUIRED KNOWLEDGE The following annunciators are symptomatic of both train RATs being de-energized.
- ALB32-D02 RESV AUX XF MR 1NXRA HI SIDE PHOC LOR TRIPTuesday, February 25, 2014 4:23:03 PM 1
- ALB32-E02 RESV AUX XF MR 1NXRB HI SIDE PHOC LOR TRIP The following annunciators are symptom atic of 1AA02 being de-energized.
- ALB35-A10 DG1A TRIP OVERSPEED
- ALB35-F10 DG1A EMERGENCY START
- ALB36-A01 4160V SWGR 1AA02 TROUBLE The following annunciators are symptomatic of 1BA03 de-energi zing and then being re-energized by the 1B DG.
- ALB37-A01 4160V SWGR 1BA03 TROUBLE alarms , then subsequently clears.
- ALB38-F10 DG1B EMERGENCY START The question stem states that the plant is in Mode 6 and 20 minute s has elapsed since the loss of powe r event occured. Per NMP-EP-110 Figure 3, an NOUE should be classified based on Loss of All Offsite Power to Essential Buses for GREATER THAN 15 minutes and one EDG is supplying the 4160VAC bus.
An upgrade to an ALERT would occur if at any time the 1B DG fails to keep 1BA03 energized.
On NMP-EP-110 Figure 2, there are two similar classificati ons. The ALERT threshold is the same as the NOUE threshold in Mode 6. The differ ence arise out of a requirement to maintain 1 trai n in Mode 6 and 2 trains in Modes 1-4. A similar NOUE also exist for a Loss of All Offsite Power to Essential Buses for GREATER THAN 15 minutes and one EDG is supplying each of the 4160VAC buses.ANSWER / DISTRACTOR ANALYSIS:A. Incorrect. Plausible. CU3 is the correc t classification. Howe ver, if the candidate does not recognize that 1BA03 is energized by the 1B DG and believes both 4160V buses are de-energized, then CA3 would
be the correct classificaton.B. Incorrect. Plausible. CU3 is the correc t classification. Howe ver, if the candidate incorrectly utilizes NMP-EP-110 Figure 2 instead of Figure 3, then SA5 would match the conditions of the stem for the, but forthe incorrect mode.
C. Incorrect. Plausible. CU3 is the correct classification. However, if the candidate incorrect diagnoses 1AA02 and be lieves it is energized by the 1A DG and also incorrect utilizes NMP-EP-110 Figure 2 instead of Figure 3, then SU1 woul d be the correct threshold.D. Correct. CU3 is the correct classifi cation. Loss of All Offsite Power to Essential Buses for GREATER THAN 15 minutes and one EDG
is supplying the 4160VAC bus in Mode 6.ANSWER / DISTRACTOR ANALYSISTuesday, February 25, 2014 4:23:03 PM 2
(5) Assessment of facility conditions and selection of appropriate proceduresduring normal, abnormal, and emergency situations.-Can the question be answered solely by knowing "systems knowledge", i.e.,how the system works, flowpath, logic, component location?
No, the answerrequires specific knowledge of emergency classification thresholds.-Can the question be answered solely by knowing immediate operator actions? No, IOAs are not addressed by this question.
-Can the question be answered solely by knowing entry conditions for AOPs orplant parameters that require direct entry to major EOPs?
No, the question does not address AOP or EOP entry conditions.-Can the question be answered solely by knowing the purpose, overall sequenceof events, or overall mitigative strategy of a procedure?
No, the answer requiresspecific knowledge of emergency classification thresholds.-Does the question re quire one or more of the following?* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or se ction of a procedure to mi tigate, recover, or with which to proceed* Knowledge of when to impl ement attachments and appendices, including how to coordinate these it ems with procedure steps* Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific s ubprocedures or emergency contingency procedures* Knowledge of administrative procedures that specify hierarchy,implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
Yes, the answer requires specific knowledge of emergency classification thresholds and determin ation of the specific classification based on current plant conditions. This is an SRO ONLYjob link associated with an SRO ONLY objecti ve. [LO-LP-40101-13 Givenan emergency scenario, and the pro cedure, classify the emergency (SROonly).]Tuesday, February 25, 2014 4:23:03 PM 3
Level: SROTier # / Group # T3 K/A# G2.4.46 Importance Rating: 4.2 / 4.2 Technical
Reference:
NMP-EP-110-GL03, Figure 2, Rev 3.0, page 122 NMP-EP-110-GL03, Figure 3, Rev 3.0, page 123 References provided:
NMP-EP-110-GL03, Figure 1, Rev 3.0, page 121 NMP-EP-110-GL03, Figure 2, Rev 3.0, page 122 NMP-EP-110-GL03, Figure 3, Rev 3.0, page 123Learning Objective: LO-TA-40002 Emergency Classification andImplementing Instructions using
NMP-EP-110 (SRO Only)
LO-LP-40101-13 Given an em ergency scenario, and the procedure, classify the emergency (SRO
only).LO-PP-11101-56 Pr edict the possibl e consequences of paralleling and load ing the EDG when a loss of offsite power is anticipated.
LO-PP-11101-30 Describe in general terms the actions that occur on a normal or em ergency start of a diesel engine up to and including the final condition of the dies el and the differences between a normal and emergency start.
LO-LP-60323-05 Given th e entire AOP, describe:
- a. Purpose of selected steps
- b. How and why t he step is being performedc. Expected response of the plant/parameter(s) for the stepLO-TA-60009A Respond to a Lo ss of Class 1E Electrical Systems per 18031-1/2LO-TA-37018 Respond to a Loss of All AC Power per 19100-CLO-TA-11021 Respond to Diesel Generator Alarms Using Procedure 17035-1/2 or 17038-1/2Question origin: MO DIFIED - HL18 NRC Qu estion # 056AG2.4.45Cognitive Level: C/A 10 CFR Part 55 Content: 43.5 Comments: You have completed the test!Tuesday, February 25, 2014 4:23:03 PM 4
- 1. 056AG2.4.45 001/1/1/LOSP- EP/C/A-4.1/4.3/NEW/HL-18 NRC/SRO/
At 10:00: - Unit 1 is in Mode 4.At 10:05 the following alarms illuminate:
- ALB32-D02, RESV AUX XFMR 1NXRA HI SIDE PHOC LOR TRIP - ALB32-E02, RESV AUX XFMR 1NXRB HI SIDE PHOC LOR TRIP - ALB35-A10, DG 1A TRIP OVERSPEED - ALB35-F10, DG1A EMERGENCY START
- ALB36-A01, 4160V SWGR 1AA02 TROUBLE
- ALB37-A01, 4160V SWGR 1BA03 TR OUBLE alarms, then subsequently clears. - ALB38-F10, DG1B EMERGENCY STARTCurrent time is 10:25:
Based on the current time, which one of the following is the correct Emergency Classification required to be declared?REFERENCE PROVIDED Alert Emergency (CA3)
Alert Emergency (SA5)
Notification of U nusual Event (SU1)
Notification of U nusual Event (CU3)
A.B.C.D.056AG2.4.45 Loss of Offsite Power Ability to prioritize and interp ret the significance of each annun ciator or alarm:
(CFR: 41.10 / 43.5 / 45.3 / 45.12) K/A MATCH ANALYSIS:
The candidate is given various alarms and indications associated with the electricaldistribution system. The candidate has to analyze the alarms to determine the plant status and determine the correct emergency cl assification, there is a time given when the event occurred that will also play into the classification.
The question is SRO only due to the Vogtle specific objective for Classification of an Emergency is an SRO only objective.Tuesday, January 21, 2014 4:27:30 PM 1
ANSWER / DISTRACTOR ANALYSIS:A. Incorrect. CA3 is a Cold Matrix classi fication, the plant is in Mode 4, not Mode 5 or 6. The mode was NOT stated in t he question but just an RCS temperature given to increase the pl ausibility the candidate may select the wrong classification matrix. If the candidate selects the wrong matrix with the given alarms, it is plausible he c ould misinterpret the event and cl assify wrong. With the multiple alarms and indication s, this can easily occur.
B. Correct. SA5 is the correct classification using t he Hot Matrix, this is still a difficult determination wi th the multiple annunciator window s illuminated. The plant is only one failure away from a total plant blackout in this condition but the candidate has to determine this and co rrelate the event has been ongoing for > 15 minutes.
C. Incorrect. SU1 is a Hot Ma trix classification. The plant is only one failure away from a total plant blackout in this condi tion but the candidate has to determine this and correlate the even t has been ongoing for
> 15 minutes. This choice is very plausible as the only difference between this and SA5 is t hat both diesels have to be carrying the buses to classify as SU1 versus 1 DG as in the correct choice.
This is a difficult determination with the mulitip le annuciators illuminated.
D. Incorrect. CA3 is a Cold Matrix classification, the plant is in Mode 4, not Mode 5 or 6. The mode was NOT stated in t he question but just an RCS temperature given to increase the plausiblity the candidate may select the wrong classification matrix. IF , the plant were in Mode 5 and the Cold Matrix required to be used, this choice would then be correct.
REFERENCES:
The following references will be provided to the candidates during the exam.
NMP-EP-110, GL03, Figur e 3, Cold Initiating Conditio n Emergency Action Level Matrix- Modes 5, 6, and Defueled Only NMP-EP-110, GL03, Figur e 2, Hot Initiating Condition Em ergency Action Level Matrix -
Modes 1, 2, 3, and 4 Only NMP-EP-110, GL03, Fi gure 1, Fission Produc t Barrier Evaluation VEGP learning objectives:LO-LP-40101-13 Given an emergency scenario, and the procedure, classify the emergency (SRO only).
This question is SRO only because the Emergency Plan is linked to a learningobjective that is specifically labeled in the lesson plan as SRO Only.You have completed the test!Tuesday, January 21, 2014 4:27:30 PM 2
NMP-EP-110-GL03 - VEGP EALs - ICs, Threshold Values and Basis Version 3.0 122 FIGURE 2 NMP-EP-110-GL03 - VEGP EALs - ICs, Threshold Values and Basis Version 4.0 124 FIGURE 3
- 1. WE11EA2.02 001/LOIT/SRO/M/F 3.4/3.9/WE11EA2.02/LO-TA-37020//HL15 NRC/
Initial conditions:
- Unit 1 experienced a LOCA.
- 19111-C, "Loss of Emerg ency Coolant Recirculation," was entered.
- RCS cooldown to co ld shutdown has been initiated. - RWST level is 8% and slowly lowering.
Current condition:
- Critical Safety Function Status Tree (CSFST) is ORANGE on Integrity.
Which one of the following completes the following statement?
The crew is required to __(1)__,
and then the Shift Supervisor
__(2)__ required to transiti on to 19241-C, "Response to Imminent Pressurized T hermal Shock Condition." (1) stop all pumps taki ng suction from the RWST (2) is (1) reduce ECCS flow from the RWST to one running train (2) is (1) stop all pumps taki ng suction from the RWST (2) is NOT (1) reduce ECCS flow from the RWST to one running train (2) is NOT A.B.C.D.K/A W/E11 Loss of Emergency Coolant Recirc. / 4 EA2.02 -Ability to determine and interpret the following as they apply to the(Loss of Emergency Co olant Recirculation):- Adherence to appropriate procedures and operation within thelimitations in the facility's license and amendmentsK/A MATCH ANALYSISThursday, February 27, 2014 9:13:31 AM 1
The question requires the candidate to make two decisions based on plant conditions while in 19111-C, "Loss of Emergency Coolant Recircul ation" - stopping all ECCS pumps and transitioning out of 19111-C. Both decisi ons challenge th e candidate's ability to adhere to the rules for using EOPs in compli ance with WOG and facilityrequirements. Actions taken in accordance with these EOPs are part of the bases for granting the facility's license.EXPLANATION OF REQUIRED KNOWLEDGE Per 19111-C steps 6 and 33 if RWST level lowers to <8%, all ECCS pumps taking suction from the RWST are to be placed in Pull-to-Lock (PTL).Transition to any ORANGE or RED CSFST wil l be made when conditions are met. Per the rules of EOP usage, CSFSTs are init iated when either step 22 of 19000-C is reached, or a transition out of 19000-C is ma de. CSFSTs remain in effect during the entire EOP network unless otherwise directed. 19111-C does NOT contain any exceptional guidance on CSFST implementation.ANSWER / DISTRACTOR ANALYSISA. Correct. The first part is correct. Per continuous actions step 6 and steps 33 and 34, if RWST lowers to <8%, all ECCS pumps taking suction from the RWST are placed in PTL.
The second part is correct. CSFS T monitoring was initiated on transition out of 19000-C. Ther e is no guidance in 19111-C that prohibits actions based on CSFSTs. Therefore, a transition to 19241-C will be made as soon as ORANGE path conditions areverified.B. Incorrect. Plausible. Th e first part is incorre ct. Per continuous actions step 6 and steps 33 and 34, if RWST lowers to <8%, all ECCS pumps taking suction from the RWST are to be placed in PTL.
However, step 15 does reduce ECCS flow to only one train. This is a mitigation strategy that prolongs RWST inventory. A candidate who does not possess the knowledge of the overall mitigating strategy of 19111-C could find it unreasonable to stop all ECCs pumps with a LOCA in progress and then transition to another procedure that does not address loss of injection flow.
The second part is correct. See the second part of choice A above.C. Incorrect. Plausible. The first part is correct. See the first part of choice A above.
The second part is incorrect.
There is no guida nce in 19111-C that prohibits actions based on CSFSTs. Th erefore, a transition to 19241-C would be made as soon as conditions for theORANGE path are verified to exist. However, step 1 of EOP
19113-C, "Recirculation Sump Bl ockage" directs the operator to "initiate monitoring CSFSTs for information only. FunctionThursday, February 27, 2014 9:13:31 AM 2
Restoration Procedures (FRP) sh ould NOT be implemented." A candidate may confuse the two EO Ps and believe transition out of 19111-C on CSFSTs is not allowed.
D. Incorrect. Plausible. The first part is in correct. See the first part of choice B above.
The second part is inco rrect. See the second part of choice C above.SRO JUSTIFICATION (10CFR43(b))(5) Assessment of facility conditions and selection of appropriate proceduresduring normal, abnormal, and emergency situations.-Can the question be answered solely by knowing "systems knowledge", i.e.,how the system works, flowpath, logic, component location?
No, CSFSTs aresymptom based, not system based.-Can the question be answered solely by knowing immediateoperator actions?
No, stopping the ECCS pump andtransitioning to FRPs is not governed by IOA's.-Can the question be answered solely by knowing entry conditions for AOPs orplant parameters that require direct entry to major EOPs?
No, the decisions made in the question require specific step knowledge.-Can the question be answered solely by knowing the purpose, overall sequenceof events, or overall mitigative strategy of a procedure?
No, the decision to transition is based on specific parameters and direction.-Does the question re quire one or more of the following?* Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or se ction of a procedure to mi tigate, recover, or with which to proceed* Knowledge of when to impl ement attachments and appendices, including how to coordinate these it ems with procedure steps* Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific subprocedures or emergencycontingency procedures.
Yes, the question requires the SRO to make a transition decision to an FRP after stopping all ECCS pumps. Thedecision requires application of EOP rules of usage, knowledge ofspecific limitations on FRP implementation, and a high level knowledge
of ECA and FRP mitigations strategy and interrelationships.* Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant norma l, abnormal, and emergency proceduresThursday, February 27, 2014 9:13:31 AM 3
Level: SROTier # / Group # T1 / G1 K/A# WE11EA2.02 Importance Rating: 3.4 / 3.9Technical
Reference:
19111-C REV 33.2, pages 6, 11, & 21References provided: None Learning Objective: LO-LP-37114-12 State the int ent of EOP 19111, Loss of Emergency Coolant Recirculation.
LO-PP-37117-04 Describe t he differences between the actions for 19113-C and 19111-C and the reason for the differences.LO-TA-37020 Respond to a Loss of Emergency Coolant Recirculation Capability per 19111-CQuestion origin: BANK - HL 15 Question # WE11EG2.4.2Cognitive Level: M/F 10 CFR Part 55 Content: 43.5 / 45.13
Comments:
Question appears to match t he KA. Transitioning to the FRG's on an Orange path is not SRO-only knowledge.
Knowledge of what to do when RWST reaches 8% may be
SRO-only knowledge. Since th e 8% RWST does not govern a procedure transition, please make sure this is required knowledge of the operator s (i.e., not minutia).- JAT 12/19/2013 (SAT)You have completed the test!Thursday, February 27, 2014 9:13:31 AM 4
Approved By Vogtle Electric Generating Plant Procedure Version C. S. Waldrup 19111-C 33.2 Effective Date ECA-1.1 LOSS OF EMERGENCY COOLANT RECIRCULATION Page Number 05/01/2013 6 of 49 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED Printed November 25, 2013 at 09:05 27 CAUTIONS If offsite power is lost a fter SI reset, action is requi red to restart the following ESF equipment if plant conditi ons require their operation: RHR Pumps SI Pumps Post-LOCA Cavity Purge Unit Containment Coolers in low speed (Started in high speed on a UV signal). ESF Chilled Water Pumps (IF CRI is reset). 4. Reset SI if necessary.
- 4. IF SI will NOT reset, THEN initiate ATTACHMENT E.
4 5. Check Containment Cooling Units - RUNNING IN LOW SPEED.
- 5. Start Cooling Units in low speed.
5 *6. Check RWST level - GREATER THAN 8%. *6. Go to Step 33.
6 7. Determine Containment Spray requirements:
7.a HV-9017A - CNMT SPRAY PMP-A RWST SUCT ISO
VLV - OPEN HV-9017B - CNMT SPRAY PMP-B RWST SUCT ISO VLV - OPEN Step 7 continued on next page
Approved By Vogtle Electric Generating Plant Procedure Version C. S. Waldrup 19111-C 33.2 Effective Date ECA-1.1 LOSS OF EMERGENCY COOLANT RECIRCULATION Page Number 05/01/2013 11 of 49 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED Printed November 25, 2013 at 09:05 27 14. Check if ECCS is in service:
- 14. Go to Step 24.
14 CCPs - ANY RUNNING.
-OR- BIT NOT ISOLATED.
-OR- RHR Pumps - ANY RUNNING IN INJECTION MODE.
- 15. Establish one train of ECCS flow:
- 15. 15 a. CCP - ONLY ONE RUNNING.
- a. Start or stop a CCP to establish only one Pump running. 15.a b. SI Pump - ONLY ONE RUNNING. b. Start or stop an SI Pump to establish only one Pump running. 15.b c. RCS pressure - LESS THAN 300 PSIG.
- c. Stop RHR Pumps.
15.c Go to Step 16.
- d. RHR Pump - ONLY ONE RUNNING. d. Start or stop an RHR Pump to establish only one Pump running. 15.d S Approved By Vogtle Electric Generating Plant Procedure Version C. S. Waldrup 19111-C 33.2 Effective Date ECA-1.1 LOSS OF EMERGENCY COOLANT RECIRCULATION Page Number 05/01/2013 21 of 49 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED Printed November 25, 2013 at 09:05 27 *30. Check if RCPs must be stopped:
- 30. 30 a. Check the following:
- a. IF neither condition satisfied, THEN go to Step 31.
30.a Seal number 1 differential pressure - LESS THAN 200 PSID.
-OR- Seal number 1 leakoff flow - LESS THAN 0.2 GPM.
- b. Stop affected RCPs.
- b. 30.b c. Close Spray Valve for idle RCP:
31 32. Check RWST level - LESS THAN 8%. 32. Return to Step 2.
32 33. Stop Pumps taking suction from RWST and place switches in PULL-TO-LOCK positions: