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TAA/MCD > TAA-MCD > Flashcards

TAA-MCD Flashcards

1
Q

Define undermoderated

A

MTC is negative

resonance escape probability is dominate over thermal utilization factor

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2
Q

Define overmoderated

A

MTC is positive

Thermal utilization factor is dominate over resonance escape probability

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3
Q

Describe core power effects of FTC

A

At higher fuel temps doppler broadening lessens per degree F

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4
Q

Why does FTC become more negative over core life

A

build in of Pu-240

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5
Q

State ECCS acceptance criteria

A 
Cladding temp <2200F
Cladding oxidation <17% of thickness
Hydrogen generation <1%
Coolable geometry
Long-term cooling
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6
Q

what is Critical Heat flux

A

the heat transfer per unit area to cause DNB

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7
Q

discuss how CHF change over core height

A

CHF decreases from the bottom to the top of the core

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8
Q

Define departure from nucleate boiling ratio

A

CHF divided by AHF at any point along a fuel rod

T.S. limit >/= 1.14

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9
Q

Discuss what the concern is with RCS hot leg becoming saturated

A

Delta-T is no longer representative of reactor power output

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10
Q

State the basis for heat flux hot channel factor

A

Fuel temp <4700F
Maintain cladding temp <2200F
Maintain DNBR >/=1.14

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11
Q

Define FqZ (Heat flux Hot Channel Factor)

A

the ratio of highest linear power density at core height Z along any fuel rod to the core average fuel rod linear power density

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12
Q

State the Kw/ft that would cause fuel melt and cladding ocidation

A

21 Kw/Ft = 4700F (fuel melt)

18 Kw/Ft = 2200F (cladding Oxidation)

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13
Q

Define FnDelta-H (Nuclear Enthalpy rise Hot Channel Factor)

A

Ratio of the total power produced by the highest power fuel rod to the total power produced by the average fuel rod power

  • is a measure of a maximum total power produced in a fuel rod
  • basis: DNBR
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14
Q

Define Axial Flux DIfference

A

AFD = Ptop - Pbottom

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15
Q

Define Axial Offset

A

Ptop-Pbottom/Ptop+Pbottom

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16
Q

State the T.S. for AFD

A

If unacceptable:
restore to within limits within 15 min
OR
restore power to <50% within 30 min

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17
Q

Summarize guidance for AFD control

A 
Inside admin band: do NOTHING
Outside band (ARO): push-pull-drift
Outside band: push-pull
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18
Q

State acceptance criteria for rod control

A

+/- 10% load change
+/- 5% per minute load ramp rate
50% step load rejection with auto steam dumps

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19
Q

State the function of the Setpoint Study

A

A setpoint study which defines NSSS control system setpoints and time constants to be used for initial plant start up and subsequent operation as verified by startup testing

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20
Q

State Reactor response to a SLOW reactivity addition accident

A

No power overshoot before RPS trip

Core protected by OTDT trip

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21
Q

State Reactor response to a FAST reactivity addition accident

A

power overshoots before rods insert
Core protected by FTC
- High Flux/ high rate trips

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22
Q

State Reactor response to a PROMPT reactivity addition accident

A

Large power overshoot
FTC limits power rise
RX eventually trips from PR flux trip

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23
Q

State why single rod withdrawal is worse than a bank withdrawal

A

A single rod withdrawal causes significant localized hot channel factor peeking

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24
Q

State RCS response to a Rod drop accident with rods in manual

A

Tave decreases adding negative reactivity

Decrease is not enough to restore power to original level

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25
Q

State plant response to a Rod drop accident with rods in AUTO

A

Rods withdraw due to PMM (if no C-11)

slight power overshoot causing rods to drive back in

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26
Q

Describe the analyzed rod ejection accident

A

LOCA and reactivity addition accident
FTC required to terminate power rise
RX trip on high flux rate
BOL full power would result in 10% fuel melt

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27
Q

During an increased Secondary Heat removal accident why is BOL with auto rods is most limiting

A

Less negative MTC
Temp and pressure are more stable
power is higher and dominant so DNBR decreases

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28
Q

State design features that minimize the consequences of steam line breaks

A 
Flow restrictor in outlet of SG nozzle
MSI actuation
SI actuation
ASME class 1 piping
Various RX trips
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29
Q

State most limiting factor in MSLB accident

A

Most reactive RCCA is assumed to be stuck fully withdrawn after reactor trip

  • allows possible re-criticality
  • results in higher peaking factors
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30
Q

Explain why HZP and EOL for MSLB is severe

A

HZP: maximum SG water inventory
EOL: MTC is most negative

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31
Q

State DNB parameters of concern

A

Tave = 595.1F
PZR press >/= 2185#
- restore within 2hrs

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32
Q

For a Decreased Secondary heat load removal accident state why a low-low SG water level with a LOP is severe

A

Minimizes the SG heat transfer capability and increase the amount of RCS stored energy at the time of RX trip

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33
Q

State how RCS Bleed and Feed is accomplished

A

Used during loss of all feed event (loss of heat sink)

- Both PORVs and one CCP + SI

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34
Q

Explain why it is important that the initiation of bleed and feed be delayed criteria is met

A 
WR level in and 3 SG < 30%
PZR pressure > 2385#
- SG mass still available for cooling
- Cooling helps repressurization
- lower RCS pressure allows more ECCS flow
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35
Q

List stages for Large Break LOCA

A

Blowdown
Refill
Reflood
Recirculation

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36
Q

Discuss SBLOCA worse case and location

A

Cold leg 4” SBLOCA is the worse case due to loop seal and that steam is prevented from being vented from the core

37
Q

State the UFSAR worse case LBLOCA

A

Loss of one train of ECCS

Both trains of CBS work

38
Q

Explain why for SBLOCAs the core can become uncovered

A

A loop seal forms in the intermediate leg which slowly depletes inventory and causes core uncovery
- once level is low enough the loop seal is broken and reflood can commence

39
Q

Describe how to determine a SGTR from a LOCA

A

an uncontrolled SG NR level increase

40
Q

Describe major actions in E-3 that are required to recover from a SGTR

A

Identify and Isolate the ruptured SG
Cooldown RCS
Depressurize RCS
Terminate SI

41
Q

Explain my stopping an RCP is not desirable during a SGTR

A

CDR for natural circ is slower

normal sprays are not available for depressurization

42
Q

For a SGTR state the UFSAR major assumptions

A

1 intact ASDV fails to open
Ruptured SG ASDV fails open
LOP
Normal plant parameters

43
Q

For a SGTR state the consequences for not meeting the TCAs

A

Water may relieve out of the SG safeties, radiological mess

44
Q

For loss of flow accidents that the core safety limit of concern

A

DNBR >/= 1.14

45
Q

Explain RCP under voltage trip reasoning

A

this is an anticipatory trip that senses flow maybe too low

46
Q

Explain RCP under frequency trip reasoning

A

The RCP flywheel won’t have enough inertia for adequate coast down

47
Q

State relationship between loss of flow and OTDT trips

A

OTDT is a DNB trip

OTDT assumes nominal flow

48
Q

Explain why a seized RCP rotor is the most severe low flow accident

A

There is no flywheel coast down

significant DNB occurs

49
Q

State the most significant post-LOCA H2 contributor

A

zirc water reaction

50
Q

State why transfer to hot leg recirc is desired

A

Prevents boron precipitation

condenses any steam in the outlet plenum

51
Q

List the 5 indications of natural circulation

A 
Subcooling >40f
Hot leg temps - stable or decreasing
Core exit thermocouples - stable or decreasing
SG pressures - stable or decreasing
Cold leg temps - at Tsat for SG pressure
52
Q

Discuss why Delta-T can not be relied upon in natural circulation

A

Delta-T is only valid for forced flow

53
Q

State limits on subcooling if natural circ is being used

A

this allows vessel head to cooldown

  • > 50F with 2 CRDMs running
  • 100 and 130F if CRDMs fans are not funning
54
Q

State how to enhance natural circ

A 
PZR level at 25%
RCS subcooling at 50f
SG NR at 50%
CDR at 50F/hr
ALL CRDM fans on
55
Q

State the two factors that determine the severity of a SBO

A

Duration of the event

Response of the RCP seals

56
Q

Discuss why the RCP seals are the most susceptible in a SBO

A

If cooling is lost then seals will degrade >550F

This will cause a LOCA and core uncovery if power is not restored

57
Q

Describe two corrective measures that must be taken to minimize a SBO

A

Cool down RCS

Depressurize the RCP

58
Q

Describe UFSAR SBO conclusions

A

4 HR duration
4 HR battery capacity
CST required volume: 137,000gallon
Core is not expected to become uncovered

59
Q

State worse case ATWS event

A

Loss of secondary heat sink

- loss of load (2959#) RCS pressure

60
Q

State assumed RX power after ATWS

A

after 10 min RX power is 5% due to steam demand and EFW capacity

61
Q

Discuss how to limit an ATWS event

A

Trip the turbine w/in 30 seconds (preserves SG inventory)

Initiate SG flow w/in 60 second

62
Q

List the operator actions to mitigate ATWS event

A

Insert negative reactivity
trip turbine w/in 30 seconds
EFW actuation within 60 seconds
Verify PORVS and safeties open when required

63
Q

State the temp for inadequate core cooling

A

CETC >1100F, assumes core damage will happen

64
Q

State conditions for inadequate core cooling

A

No CCPs or SIPs

65
Q

State in order of preference inadequate core cooling recovery methods

A

ECCS actuation
Secondary Depressurization
RCP restart
RCS depressurization

66
Q

Explain response of normal response of NI following a reactor trip

A

Prompt drop to 6%
stable -1/3 SUR
SR detectors should reenergize at 5e-11 about 15 to 20 min

67
Q

Explain what determines equilibrium fission rate in a shut down RX

A

Directly proportional to source strength (higher at EOL)

Inversely proportional to SDM

68
Q

Why does SR counts increase as voiding increases with RCPs running

A

RCS is saturated and progressively voiding
Indicated SR level rises as a much larger fraction of the neutron population reaches the NIs (less shielding from coolant)

69
Q

Why does SR counts change as voiding increases without RCPs

A

SR counts increases as core uncovers and downcomer empties

SR counts then decrease as level drops further and source strength drops

70
Q

Explain the response of incore NI as the core is uncoverd

A

During uncovery the outputs above the water level rise significantly
The outputs below the water level will read lower and show more variation

71
Q

State how to calculate Subcooling margin

A

Use WR RCS pressure (PT-403,405)
Determine Tsat for pressure
Subtract highest average CETC quadrant temperature

72
Q

What does a negative subcooling margin mean

A

Indication of superheat at core exit

73
Q

What does 40F subcooling margin mean

A

Saturation temperatures at core exit

74
Q

Discuss why CDR is more restrictive

A

CDR is more restrictive due to the tensile stress on the inner vessel wall causing the vessel to be closer to max allowed stress

75
Q

State the four conditions that must be present for Pressurized thermal shock

A

Severe cooling and high rate
High internal pressure
High NDT
Critical flaw

76
Q

State the three things as an operator that can be done to reduce PTS

A

Stop the cooldown
Terminate SI
Soak the RCS

77
Q

State the Containment design basis for a LBLOCA

A

LBLOCA in the intermediate leg

Combines effect of high core reflood rate plus SG heat addition

78
Q

State issue with FR-Z.1 versus other FRPs

A

FR-Z.1 there is no alternate recovery actions can be provided because CBS is the only ESF system that will mitigate a rise in containment pressure

79
Q

State parameters for entry into Containment status tree

A

pressure >52#
OR
Pressure >18# and Phase A and B isolations not isolated

80
Q

List the four major classifications on post accident fuel conditions

A

Undamaged fuel
Failed cladding
Overheated fuel
Core Melt

81
Q

Describe the fission products unique to each type of post accident fuel conditions

A

Failed cladding - Xenon, Krypton, Iodine
Overheated Fuel - Xenon, Krypton, Iodine and CESIUM
Core Melt - Strontium and Barium

82
Q

State auxiliary parameters to asses core damage

A

Containment H2
CETCs
RVLIS
Containment radiation level

83
Q

Describe how Iodine spiking occurs

A

Iodine transfuses through a cladding defect into the RCS

84
Q

Describe where Post accident samples can be drawn from

A 
Chemistry samples from Loops 1 or 3
OR 
RHR pump discharge
all through the PASS System
Radiation on sample can be very high
85
Q

State which Radiation monitor is most useful to the plant to determine containment rad levels

A

Post LOCA Containment area radiation monitors

86
Q

State how to estimate radiation doses in containment

A

Take external dose rates at the hatch and solve for internal dose

87
Q

Describe how the operator determines the dose to the public

A

Raddose V is used to calculate

88
Q

State what must be done to use CSF for EAL entry

A

Conditions must be evaluated and verified via hardwire indications

TAA/MCD (1 decks)

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