TA-MCD Flashcards
(152 cards)
1
Q
What are the 2 major effects of a Radiological Accident?
A
- Surface contamination
a) Agriculture, food chain, pollution - Personal Injury
a) Radiation injuries
2
Q
What is the major threat to the population after a reactor accident?
A
- Radioactive fallout from the discharge flume
3
Q
The extent to which the contamination spreads depends upon which factors?
A
Amount released, height of release, wind speed and direction Core Damage
- Fission product activity that exceeds TS
- Fuel is no longer in original geometry
- Major portion of the core cannot be operated for its design cycle length
4
Q
What is the most likely event to cause core damage?
A
Inadequate Core Cooling
LOSP
5
Q
Describe core damage threshold
A
- Onset of gross cladding failure and core geometry
6
Q
Will core damage occur if safety limits are not exceeded?
A
- If safety limits are not exceeded, core damage should not occur
7
Q
Definition of Safety Limits
A
a) Limits to ensure fuel design limits are not exceeded
8
Q
Definition of Limiting Safety System Settings (LSSS)
A
a) Limiting safety System Settings
(1) Automatic action setpoints
(2) RPS for fuel clad and reactor pressure boundaries
9
Q
Definition of Limiting Condition of Operation (LCO)
A
a) Limiting Condition for Operation
b) Minimum acceptable levels of system performance necessary to ensure safe startup and operation of the plant
c) RAS will be required action to ensure safety if LCO is violated
10
Q
Radioactive material barriers to prevent release
A
- Fuel pellet
- Cladding
- Moderator
- RPV / Piping
- Primary Containment
- Secondary Containment
11
Q
Safety Limits and Basis
A
- <685# or <10% flow, maintain <24% power
a) 14.7 to 800# crit power at minimum flow >50% power
b) Margin built in - > 685# and >10%, maintain MCPR>1.07
a) 95% probability at 95% confidence transition boiling will not occur - RWL shall be > the TOAF (155”)
a) Core covered to remove decay heat - <1325# in the steam dome
a) 1250# + 10% = 1375#
b) 1325# at the dome ensures bottom of head is <1375#
12
Q
Basis for Power Distribution LCO
A
- Normal operation and abnormal transients to maintain fuel cladding integrity (expected parameter change)
- Postulated accidents to maintain core geometry
- LHGR
- MCPR
13
Q
Probilistic Risk Assessment Purpose
A
- Realistic evaluation are most likely to cause core damage, and how often
a) Hazards
b) Initiating Events
c) Frequency
d) Probability of failure
e) Core damage freq, large early release freq, and Offsite Dose Consequence
14
Q
Five most probable core damage events listed in the PRA
A
- LOSP – equipment failure (all others are operator error)
- Loss of PSW
- Loss of DC Bus
- MSIV Closure
- Turbine Trip
15
Q
What is the Worst Case Scenario regarding core damage scenarios?
A
- Loss of RPV Makeup
- Loss of DHR
- ATWS
16
Q
Mitigating risk is a __________
A
J. Mitigating Risk is a 10 CFR “LEGAL requirement”
17
Q
Chernobyl Accident
A. Describe the test being performed
A
- To determine the length of time that the turbine would supply power at near rated voltage on a turbine trip
- # 8 TG was to power 4 recirc pumps and 2 feed pumps
- EDG would come on at 3 minutes
18
Q
Chernobyl Accident
B. Describe the sequence of events
A
- 0100 – power reduction to 700MW
- 13:05 - #8 TG is supplying required pumps for test
- 14:00 – ECCS systems disconnected per test
- The power reduction was paused due to load dispatcher
- Night shift turnover
- 23:10 – Downpower resumed
- 01:00 – power is too low because Xe, withdrew to 6/8 rods in (req 30)
- 01:03 – added recirc pumps to raise power, exceeded max flow – Tsat
- 01:19 – can’t control pressure and level, bypassing auto SCRAMs, shut TBV
- 01:22 – rods are less than 6, no manual or auto SCRAM
- 01:23 – initiate test, TSV shut, pressure goes up, flow does down, more voids, more reactivity, more explosion
- 320000MWth (100 times normal) – core geometry damaged, no rods in
- Manual scram caused rupturing fuel tubes and thermal explosion
19
Q
Chernobyl Accident
C. Describe the procedural violations
A
- Reduced power <22% (unstable)
- Reduced power below test requirements
- Too much recirc flow
- Overrode SCRAM functions
- Overrode ECCS
20
Q
Chernobyl Accident
D. Describe lack of management controls and inadequate safety review
A
- No oversight
- Test conducted by junior electrical engineer
- Not adequate safety review
21
Q
Chernobyl Accident
E. Describe the consequences
A
- 31 died immediately
- 29/35 died of rapid responders
- 4.3 miles of the plant was 54 REM
22
Q
TMI Accident
A. Basic Events that lead to a loss of all feedwater
A
- Misposition of Emergency Feed Water block valve
2. Operator error prevented 2nd High Pressure Makeup pump
23
Q
TMI Accident
B. Reason why the operators at TMI believed the ERV closed after the initial pressure transient
A
- Indication of ERV is based on signal to open the valve
- Same as at Hatch, not based on valve position
- Pressure was low enough to have valve reseat
- Masked tailpipe high temperature due to known leakby
24
Q
TMI Accident
C. Describe the release path from the core
A
- 200gpm from the ERV – about 1” a minute at Hatch
- Reactor Coolant Drain Tank to Aux Building Waste Tank
- ABWT overflows to AB Sump, and release to environment
25
TMI Accident
| D. What plant conditions caused the reactor coolant pumps to vibrate
1. Cavitation due to saturation conditions
26
TMI Accident
| E. What caused the operators to believe the reactor was supercritical
1. Voids in the core caused SRNI to read high
2. Chemistry sample of low boron concentration
3. Diluted by condensing steam therefore not indicative of actual coolant
27
TMI Accident
| F. What made the operators to finally realize the ERV was open and a LOCA was occurring
1. Identified by a new arrival into the CR
28
TMI Accident
| G. Discuss the conditions surrounding the hydrogen burn which occurred inside the primary containment
1. ERV block valve opened to reduce pressure, releasing H2 to RB
2. 28# pressure spike due to burn
29
TMI Accident
| H. Basic consequences of TMI, including the amount of core damage, area evacuation, and radiation exposures to personnel
1. Melted UO2 at 5100F
2. 30000 of 37000 fuel rods shattered or melted
3. 450kg of H2 produced via Zircaloy reaction
4. 100,000,000 curies released from fuel
5. 10,000,000 curies released to environment
6. 20 curies of Iodine released
7. Highest offsite dosimeter read 277 mrem
8. 4200 pregnant women and children mandatory evac
9. 144000 people evacuated themselves
30
TMI Accident
| I. Describe the behavior of the incore and excore instrumentation systems during the accident
1. Incore temperatures had to be read by voltmeter due to offscale
2. Excore Source Range detectors responded to voids as uppowers
31
TMI Accident
| J. Describe the four fundamental causes of the accident at TMI
1. Equipment operated in degraded conditions for extended periods of time
2. Plant indications were discounted and considered faulty, leading to misrepresenting of symptoms and not taking corrective actions
3. Event Based EOPs and Abnormals
4. Training and Qualification of operators was inadequate
32
TMI / Chernobyl Lessons Learned
A. Given a description of the TMI and Chernobyl accidents, Discuss the areas where plant and performance improvements were made based on lessons learned.
1. TMI
a) Mitigating Core Damage Training
b) Thermodynamics Training
c) Natural Circulation Principles of Operation
d) Effects of rapid pressure drop on systems operating at saturation or slightly subcooled.
e) Drywell temperature may have an adverse effect on heated reference leg level instruments
f) Effects of a throttling process across a valve
g) Common reference level for RPV level
h) Redundant instrumentation and uses
i) Necessity of operators to systematically analyze plant conditions
j) Do not make operational decisions based on a single plant parameter
k) Don’t dismiss a parameter just because it is not clearly understood
l) Don’t dismiss nuisance alarms
m) STA
n) Don’t override ESF system unless you are sure that the ESF will cause unsafe condition
o) Don’t secure an automatic function unless it is clearly understood that it is not operating properly
p) Don’t stop both trains of a safety system SCARY!!
q) The need for redundant containment pressure indication in the control room
r) Hydrogen Control should be minimized by inert containment
s) Diverse containment isolation
t) Airborne iodine concentrations within the facility must have equipment, training and procedures
u) High range noble gas effluent monitors recommended
v) Human factors design review of control rooms were conducted
w) Shift manning changes and hour restrictions
x) SPDS
y) Emergency Plans should be upgraded to include radiological events
z) Establishment of the TSC
aa) Establishment of the EOF
33
TMI / Chernobyl Lessons Learned
A. Given a description of the TMI and Chernobyl accidents, Discuss the areas where plant and performance improvements were made based on lessons learned.
2. Chernobyl
a) Overall management control of plant evolutions must be maintained
b) Procedures for the conduct of special tests, including test procedures, must be reviewed for effects on nuclear safety and approved
c) Tests should include initial conditions, predictions of how the test should proceed and respond, acceptance criteria, test termination and abandonment criteria, provisions to returning the plant to safe operating conditions, and safety analysis as applicable
d) Operators should be fully briefed, or else terminated
e) Strict adherence to safety requirements
f) Operator training on reactor behavior including radioactivity effects and transient and accident analysis. Understand safety consequences of improperly operating control rods, bypassing safety circuits, and operation in abnormal configurations
g) Misoperation of reactivity controls can place plant in unanalyzed condition
h) Mitigating Core Damage Training for assessing and containing a damaged reactor
i) Loss of containment resulted in contamination spread across the world
j) Maintain Containment Integrity!
k) PR – use measurements and units that the public will recognize
34
TMI / Chernobyl Lessons Learned
A. Given a description of the TMI and Chernobyl accidents, Discuss the areas where plant and performance improvements were made based on lessons learned.
3. Industry Chages
a) Increased communication between Nuclear Utilities
b) Davis Besse very similair event to TMI 1.5 years prior
c) Transient did not damage fuel or equipment
d) Lessons learned were not shared with other utilities
e) Had the TMI operators been informed of the ERV failing open, TMI might have been relatively minor transient.
f) Many different forms of communication now commonplace (SOERs, NRC Bulletins)
g) At time of event procedures were event oriented. Operators had to recognize the transient and take action from memory
h) Procedures changed to be symptom oriented. Operators now treat the symptoms of the transient.
i) Early power plant designs concentrated on safety related equipment combating accidents
j) After the TMI accident, more emphasis is placed on the role of the operator in nuclear safety.
k) Indications available in control rooms are improved to assist the operator in recognizing “symptoms” of transient.
l) Much more simulator training is now required to obtain and maintain a license to operate a Reactor.
m) Topics such as “Team Training” and “Transient Analysis” are now seen in operator training.
n) An extensive task analysis has also been performed for most nuclear plant personnel by INPO.
o) Training for plant personnel is now task “oriented.”
p) NUREG - 0737 contains many changes in the design of power plants such as better Containment Isolation dependability, Post Accident Sampling, accident monitoring and control room habitability systems.
q) Since TMI accident, STA has provided an extra set of eyes to SS.
r) The STA independently assesses plant conditions and advises the SM/SS
s) TMI and Chernobyl were not the only accidents to happen in the nuclear industry but both events taught valuable lessons.
t) TMI pointed out many inadequacies in the industry, and as a result, the utilities made numerous changes to insure an event like this does not happen again.
u) Training improved, especially in plant fundamentals and AB’s and EOPs.
v) Installing belief in trusting plant indications
w) Understanding NI response under accident conditions.
x) ECCS systems should be allowed to perform function unless absolutely certain will cause unsafe plant condition
y) Containment radiation and hydrogen detection monitors were upgraded to indicate higher concentrations that occur during accident conditions.
z) Procedures were changed from event based to symptom based so that the exact cause of the event does not have to be known to start to combat the situation
aa) Overall management control of plant evolutions must be established and maintained.
bb) Safety measures and controls must be in place and should not be overridden without careful consideration and review and only with management approval
cc) Test procedures should include initial conditions, predictions of how the test should proceed and expected response, acceptance criteria, test termination and abandonment criteria, provisions for returning the plant to normal operating or shutdown conditions, and safety analysis as appropriate.
dd) Misoperation of reactivity controls can place the plant in an unanalyzed condition, compromising safety, as well as being an outright violation of procedure
35
Fukushima Daiichi
| A. Describe the event that triggered the SCRAMS
EARTHQUAKE!
1. RPS scrams on seismic events
2. 9.0 Richter Scale – 112 miles off the coast
36
Fukushima Daiichi
| B. Describe the event that triggered LOSP
EARTHQUAKE!
1. Seismic Acceleration instrument
2. Tsunami caused loss of emergency power
3. Exceeded height and horizonal “g-force” design bases
37
Fukushima Daiichi
| C. Describe Reactor Pressure Control systems operated during Extended Station Blackout
1. U1 - Isolation Condenser was the only form of pressure control
2. U2 – Relief Valves and RCIC
3. U3 – Relief Valves and RCIC
38
Fukushima Daiichi
| D. Worst case effects of a prolonged lack of DHR on cladding
1. Zirconium Water reaction creates Hydrogen Gas - explosion
39
Fukushima Daiichi
E. State the threat to the Reactor Building that can be posed by cladding oxidation compounded by excessive Containment Pressure
1. Explosion from H2 accumulation
40
Fukushima Daiichi
| F. Highest Total Radiation Dose received by an operator
1. 67.8 REM
41
Recognizing Core Damage
| A. DESCRIBE the term core damage.
1. Failure of the fuel clad integrity to the extent that any of the following conditions exist:
a) Fission product activity in coolant > TS
b) Fuel no longer in original geometry
c) Major portion of the core cannot be operated for entire cycle
42
****IMPORTANT****
Recognizing Core Damage
B. DESCRIBE the three basic mechanisms which could cause core damage in a BWR.
* ***IMPORTANT****
1. Severe overheating due to inadequate core cooling
2. Rupture due to strain
3. Rupture due to overpressure due to FP gas buildup
43
Recognizing Core Damage
| C. DESCRIBE what two factors determine how much of the fuel cladding reacts with the steam during a LOCA.
1. Length of time clad uncovered
| 2. Heat available for reaction
44
Recognizing Core Damage
| D. DESCRIBE the goal of the Emergency Core Cooling Systems following a LOCA.
1. Plant designed to handle Worst Case LOCA without operator action until after 10 minutes
2. If ECCS works properly, the cladding will remain intact to retain pellets for easily coolable array following a LOCA
3. Peak cladding Temp of 2200F and max oxidation of 17%
45
Recognizing Core Damage
E. DESCRIBE why the potential for fuel failure exists as the Core Spray system attempts to reflood the core following a large LOCA.
1. Assume CS delayed, Clad temps rise until 2/3 of core in dryout, then CS injects
2. Removes Radiation Heat to steam, Radiation Heat to water droplets, Convection Heat on water droplet impingement
3. Shattering of the clad as the quench front moves down the core
46
Recognizing Core Damage
F. DESCRIBE how the fuel cladding can rupture during a LOCA even if cladding temperature is maintained well below 2200°F.
1. Four Major Release Mechanisms
a) Gap Release
b) Meltdown Release
c) Oxidation Release
d) Vaporization Release
2. Rise in Ductility
a) Cladding can balloon and Rupture
47
Recognizing Core Damage
| G. DESCRIBE how activation products are introduced in the reactor coolant during normal operations.
1. Oxygen produced N-16
| 2. Corrosion products
48
Recognizing Core Damage
| H. DESCRIBE how fission products are introduced into the reactor coolant during normal operations.
1. Direct Recoil
a) Uranium within .003 cm of coolant channel
2. Cladding effects
49
****IMPORTANT****
Recognizing Core Damage
I. DESCRIBE the following fission product release mechanisms during a core damage event.
1. a. Gap Release
* ***IMPORTANT****
a) Gases in the fuel pin through hole or rupture
b) pressure in the fuel pin > Rx Pressure
c) Release rate goes up as Temp goes up
50
****IMPORTANT****
Recognizing Core Damage
I. DESCRIBE the following fission product release mechanisms during a core damage event.
2. b. Meltdown Release
* ***IMPORTANT****
a) Gases in the pellets are released as fuel melts
b) Can bubble up through molten core, arriving in upper internals
51
****IMPORTANT****
Recognizing Core Damage
I. DESCRIBE the following fission product release mechanisms during a core damage event.
3. c. Oxidation Release
* ***IMPORTANT****
a) Steam explosions when molten core hits water at bottom of vessel
b) Aerosol fuel scattered in atmosphere causes excessive oxidation
c) Liberates a ton of FP gases – may rupture vessel
52
****IMPORTANT****
Recognizing Core Damage
I. DESCRIBE the following fission product release mechanisms during a core damage event.
4. d. Vaporization Release
* ***IMPORTANT****
a) Molten Core reacts with concrete at bottom of Drywell
b) Gases produced through explosions after RPV melts
53
Recognizing Core Damage
| J. DESCRIBE the general conditions inside the core during meltdowns.
1. Steam - Hydrogen mixture with FP / Core material vapors and aerosols
2. Highest powered portions of the core have melted and dripped into the lower portions of the core
3. Some fuel may still be intact and needs cooling
54
****IMPORTANT****
Recognizing Core Damage
K. DESCRIBE how the operator can use the Neutron Monitoring System and the Control Rods to obtain an indirect indication of gross core damage.
* ***IMPORTANT****
1. Inability to insert detectors
2. Inability to insert control rods
3. Radiation levels rise in reactor building
55
Core Cooling and Reactivity Control
| A. DESCRIBE adequate core cooling.
1. Heat removal from the reactor sufficient to prevent rupturing the fuel load
56
****IMPORTANT****
Core Cooling and Reactivity Control
B. DESCRIBE four acceptable methods of assuring Adequate Core Cooling.
* ***IMPORTANT****
1. Core submergence
2. Steam cooling with injection
3. Steam cooling without injection
4. Spray Cooling
a) 4250gpm when >-207”
57
Core Cooling and Reactivity Control
| C. DESCRIBE how to verify core submergence.
1. RWL > TAF = -155”
| 2. RPV flooding has been determined in the Main Steam Lines
58
Core Cooling and Reactivity Control
| D. DESCRIBE how to verify proper steam cooling with injection.
1. RWL > -180”
| 2. <1500F or Emergency Depressurize Vessel
59
****IMPORTANT****
Core Cooling and Reactivity Control
E. DESCRIBE how to verify proper steam cooling without injection.
* ***IMPORTANT****
- 195" is the minimum zero injection RWL
1. RWL > -195”
2. <1800F or ED the RPV
60
Core Cooling and Reactivity Control
| F. DESCRIBE the sources of heat found in a reactor after shutdown.
1. Sensible Heat
a) Heat stored in coolant, vessel walls, internals
2. Decay Heat
a) Generated by delayed neutron fission and the decay of fission products
b) After 1 second, normal cooling has removed 80% of decay heat
c) Total decay heat in the core 10 seconds after a SCRAM is 6%
61
Core Cooling and Reactivity Control
G. DESCRIBE how the following modes of heat transfer can be used to remove heat from a damaged core:
1. a. Conduction
a) Surface Area
b) Thermal Conductivity
c) Thermal Gradient across the material
62
Core Cooling and Reactivity Control
G. DESCRIBE how the following modes of heat transfer can be used to remove heat from a damaged core:
2. b. Convection
a) Natural
| b) Forced
63
Core Cooling and Reactivity Control
G. DESCRIBE how the following modes of heat transfer can be used to remove heat from a damaged core:
3. c. Radiation
a) In severely damaged core only when >2000F
64
Core Cooling and Reactivity Control
| H. DESCRIBE how fuel cladding failure can affect core cooling.
1. Ballooning or Rupturing of fuel causes restrictions
2. Perforations most likely midplane for fuel damage
3. Central region of the core has lowest flow
65
Core Cooling and Reactivity Control
| I. DESCRIBE the effect of break size on the long-term core cooling after a LOCA.
1. Small to Medium
a) RPV will reflood to cool the cladding
2. Large
a) RPV will reflood to 2/3 core height (jet pump suction)
b) Lower part of the core is cooled by submergence, and upper part by steam cooling
66
Core Cooling and Reactivity Control
| J. LIST the parameters that will affect the reactivity in a shutdown reactor.
1. Moderator Temp
2. Fuel Temp
3. Void Fraction
4. Poison Concentration
5. Control Rod Density
6. Core Mass (fuel concentration)
7. Core Geometry (core shape)
67
Core Cooling and Reactivity Control
K. DESCRIBE how changes in various plant parameters would affect the concentration of sodium pentaborate in a damaged core.
1. Boron dilutes during injection
2. Boron leaves due to LOCA
3. Boiling Rate / ECCS flooding Rate / Core Temperature
68
Use of Radiochemistry to Determine the Extent of Core Damage
| A. DESCRIBE volatility and its effect on the transport of radionuclides during an accident.
1. Volatility is primary transport mechanism of radionuclides
2. Vaporize at low temperatures
69
Use of Radiochemistry to Determine the Extent of Core Damage
| B. STATE the effect of the water in the torus on the release of radionuclides during an accident.
1. SRVs discharge under the Torus water level, which “Scrubs” the release
70
Use of Radiochemistry to Determine the Extent of Core Damage
C. DESCRIBE the removal mechanisms for the following categories of radionuclides:
1. Iodine
a) Trapped in pellet-clad gap, released as Cesium-Iodine
b) Will be “scrubbed” via Torus
c) Trapped in Condensation on cooler vessel and drywell walls “plating out”
71
Use of Radiochemistry to Determine the Extent of Core Damage
C. DESCRIBE the removal mechanisms for the following categories of radionuclides:
2. Aerosol particles
a) Settling
b) Filtration via SBGT
c) Torus Scrubbing
72
Use of Radiochemistry to Determine the Extent of Core Damage
C. DESCRIBE the removal mechanisms for the following categories of radionuclides:
3. Noble gas
a) Activated Charcoal
73
Use of Radiochemistry to Determine the Extent of Core Damage
| D. DESCRIBE the fission product source term during an accident.
1. Fraction of the radioactivity inside the fuel rod which is released to the coolant
74
Use of Radiochemistry to Determine the Extent of Core Damage
E. DESCRIBE the basic steps performed when using the Post Accident Sampling System to determine the extent of core damage.
1. Grab sample from Reactor Coolant and Drywell atmosphere
2. Gamma isotopic analysis to determine concentrations
a) Coolant – I and Cs
b) Drywell atmosphere – Xe and Kr
3. Correction factors for time since shutdown, sample dilution, power history and temp/press changes
75
Use of Radiochemistry to Determine the Extent of Core Damage
F. DESCRIBE the basic steps performed when using the Drywell Wide Range Radiation Monitors to determine the extent of core damage.
1. CR provides TSC Drywell Radiation from D11-K621A/B
a) These monitors also supply the 138R/hr Fast Vent Closure trip
2. Correction factors applied to compensate for drywell volume and distance
76
Use of Radiochemistry to Determine the Extent of Core Damage
G. DESCRIBE the basic steps performed when using the Drywell Hydrogen Concentration to determine the extent of core damage.
1. H2O2 analyzers in service after a Group2
2. Correction factor to give a readout in Zr% in steam
3. Correlates to fuel damage
77
Gas Generation during Degraded Conditions
| A. Describe the five major sources of H2 during an accident
1. Zirconium Water Reaction – primary method
2. Steel Steam reaction
3. Concrete Decomposition
4. Radiolysis of Water
5. Corrosion of zinc based paint, galvanized steel and aluminum
78
Gas Generation during Degraded Conditions
| B. Two items required for Zirc/Water reaction
1. Steam
| 2. Zirconium at high temps from fuel cladding
79
Gas Generation during Degraded Conditions
| C. Explain how Zirc/Water reaction occurs and how it is affected by rising temps
1. Zr + 2H20 = ZrO2 + 2H2 + Energy
2. Four steps
a) Steam diffuses through the hydrogen layer which will form on the oxidized surface of the cladding
b) Water molecule disassociates
c) O2 diffuses through oxide layer into the zirc
d) Allows unreacted zirc to interact with O2, making ZrO2
(1) Oxidizes rest of cladding and creates a ton more H2
80
Gas Generation during Degraded Conditions
| D. Explain how steel components in upper vessel and lower plenum areas could be heated to cause oxidation
1. Steam heats it during steam cooling
81
Gas Generation during Degraded Conditions
| E. Describe what core conditions must exist before decomposition of concrete can produce H2
1. Molten core at the bottom of the drywell
2. Three stages
a) Thermal composition of concrete produces gases
b) Gases go up through corium produces flammable gases
c) Chemical evolution of gases as they escape the corium (H2, CO)
82
Gas Generation during Degraded Conditions
| F. Discuss radiolysis of water
1. Neutron Hammer
83
Gas Generation during Degraded Conditions
| G. Discuss how H2 by radiolytic decomposition of water varies with changes in decay heat
1. As Decay heat goes away, H2 production by radiolysis will go away
84
Gas Generation during Degraded Conditions
| H. Describe Two major sources of zinc inside the drywell
1. Paint
| 2. Galvanizing coat
85
Gas Generation during Degraded Conditions
| I. Discuss the strengths of FIVE major H2 production reactions
Ordered from Greatest to Least
1. The zirc/steam reaction (most dominant)
2. Steel/steam
3. Concrete decomposition.
4. Radiolysis
5. The corrosion of paint and other surfaces within the primary containment
86
Gas Generation during Degraded Conditions
| J. Explain how H2 can escape from the reactor into primary containment
1. LOCA
| 2. SRV operation if MSIVs are closed
87
Gas Generation during Degraded Conditions
| K. How H2 mixes in the primary containment atmosphere
1. Differential pressure
2. Natural convection
3. Forced convection
4. Diffusion (H2 - gases fill their container in equal concentration)
88
Gas Generation during Degraded Conditions
| L. State upper and lower flammability limits of H2 in the atmosphere
1. 4% to 76% is flammable
| 2. 4-18% and 58-76% are susceptible to deflagration
89
Gas Generation during Degraded Conditions
| M. Explain how inerting drywell with N2 affects H2 flammability
1. Lowers upper limit rapidly and raises lower limit
| 2. When limits meet, no flammable region exists anymore - inerted
90
Gas Generation during Degraded Conditions
| N. Deflagration of H2, including speed of combustion wave
Deflagration is a self-sustaining rapid burn with intense heat. The resulting combustion waves, which normally travel subsonically (2 to 50 ft/s), propagate the burn by conducting heat from the hot burned gas to the cooler unburned gas.
91
Gas Generation during Degraded Conditions
| O. Turbulence associated with containment sprays or drywell cooling fans
1. Completes combustion
92
Gas Generation during Degraded Conditions
| P. how the turbulence associated with containment sprays or drywell cooling fans affects speed of combustion wave
1. turbulent flames mean speed is 2-5 times faster with forced circulation
93
**IMPORTANT**
Gas Generation during Degraded Conditions
Q. Detonation of H2
* *IMPORTANT**
1. Detonation is a very rapid self-sustaining burn. The resulting combustion waves travel at SUPERSONIC speeds, compressing the unburned gas and heating it up to its auto ignition temperature, which causes an immediate reaction.
2. Vigorous shock wave from an explosion or a very large spark required to initiate a detonation
3. 6600ft/sec
94
Gas Generation during Degraded Conditions
| R. How presence of steam can affect the combustion of H2 and air mixture
1. Steam dilutes the mixture and inhibits combustion
| 2. <6% H2, or <5% O2, or >60% steam is SAFE
95
Gas Generation during Degraded Conditions
| S. Initiating drywell cooling could affect combustion of a hydrogen-steam-air mixture
Drywell cooling would lower the concentration of steam in the drywell. This in turn would result in a relative higher concentration of hydrogen and oxygen, and possibly cause the mixture to become combustible
96
Gas Generation during Degraded Conditions
| T. Effect of drywell size and shape on the combustion
1. The bigger the drywell and more free space the better
97
Gas Generation during Degraded Conditions
| U. Different methods which could be used to control the H2 concentration
1. Inert the drywell with N2 to Limit O2
2. Purge Containment with feed and bleed using N2 or Air in and H2 through SBGT out
3. A combination of any of the above items.
98
Release Pathways and Radiation Monitoring during Core Damage
| A. DISCUSS why gaseous and airborne particulate releases are more likely to occur than liquid releases.
1. Gaseous and airborne particulate radionuclides
a) More likely as gas can readily escape by venting or leakage, or coming out of solution
2. Radionuclides in liquid
a) More easily contained
99
Release Pathways and Radiation Monitoring during Core Damage
| B. STATE which areas of the plant may become high radiation areas during severe accident conditions.
1. Core Spray / RHR diagonals during operation
2. HPCI room during operation
3. Torus area during SRV lifting
4. 158’ elevation of RB near CS lines
5. SRVs lifting to control pressure
a) Fission products directed to Torus
b) Torus Area Monitors and RB monitors due to “Shine”
100
****IMPORTANT****
Release Pathways and Radiation Monitoring during Core Damage
C. DISCUSS why special precautions should be taken while sampling the reactor coolant after a severe reactor accident.
****IMPORTANT****
1. Special RWP to draw sample during accident conditions
2. Much higher than normal radiation levels
3. 1ml could give Dose Rate of 8541 Rem/hr at 1cm
101
Release Pathways and Radiation Monitoring during Core Damage
D. STATE the purposes of the following plant radiation monitoring systems:
1. Process Radiation Monitoring System
a) Monitor radiation levels in process streams
(1) Normal release paths
(2) Potential accident release paths
b) Provides
(1) Readouts
(2) Alarms
(3) Protective actions to minimize possibility of exceeding established release limits in some cases
102
****IMPORTANT****
Release Pathways and Radiation Monitoring during Core Damage
D. STATE the purposes of the following plant radiation monitoring systems:
2. Area Radiation Monitoring System
****IMPORTANT****
a) Provide plant operating personnel with assurance that rad levels are low enough to work
b) Non-safety system designed to function during:
(1) Normal plant operating
(2) Abnormal transient conditions
(3) Should survive a LOCA because they are outside primary containment
103
Release Pathways and Radiation Monitoring during Core Damage
D. STATE the purposes of the following plant radiation monitoring systems:
3. Containment Atmosphere Monitoring System (CAMS)
a) Safety system for after core damage
b) Monitor:
(1) Drywell FPs
(2) Gammas in Drywell and Torus
(3) H2 and O2 in Primary Containment
(4) Air Temp in Primary Containment
(5) Pressure in Drywell and Torus
(6) Torus Water Temperature
104
Release Pathways and Radiation Monitoring during Core Damage
| E. DISCUSS the operation of a GM tube when placed in very high radiation fields.
1. Can Saturate
a) So high, recombination does not occur = zero output
b) Newer instruments will fail high, all other failures are low
105
Release Pathways and Radiation Monitoring during Core Damage
| F. DISCUSS the three major failure modes of a scintillation detector.
1. Phosphorescence phenomenon
a) Temperature rises gives higher electron energy levels
b) Results in a delayed photon emission which builds up to high levels
c) Will read higher than it normally is
2. Saturation
a) One microsecond resolving time
b) Downscale reading because most types don’t have fail high function
3. Crystal airtight container fails
a) Moisture absorption of crystals causing deterioration
b) Lowered detector sensitivity
106
Instrument Response during a Casualty
| A. DISCUSS how RPV water level is measured using differential pressure.
1. Fluid columns result in DP
2. Aux chamber in a Yarway
a) When reference leg flashes due to pressure drop, aux chamber empties into condensing chamber, refilling ref leg through aux chamber port
b) This colder water inhibits flashing and normalizing level
c) Steam can enter top of aux chamber through aux port and equalizing tube, refilling aux chamber
107
Instrument Response during a Casualty
| B. STATE the four ranges of level indication displayed in the control room.
1. Floodup 0 to 400”, 0 to 200”
2. Normal Control Range (Narrow) 0 to 60”
3. Wide Range -150 to 60”
4. Fuel Zone -317 to -17”
108
Instrument Response during a Casualty
C. DESCRIBE the following RPV water level instruments, including their indication range, the location of indicators and recorders, calibration conditions, and possible use during post accident conditions.
1. Narrow Range Instruments
2. Wide Range Instruments
3. Floodup Range Instruments
4. Fuel Zone Range Instruments
a) Study the Water Level Correction AB in folder – 34AB-B21-002
109
Instrument Response during a Casualty
D. DISCUSS how the following factors could cause indication errors on the RPV water level instruments.
1. A change in variable leg temperature
a) Density Error
| b) Tup, Lind down
110
Instrument Response during a Casualty
D. DISCUSS how the following factors could cause indication errors on the RPV water level instruments.
2. Two phase flow inside the reactor vessel
a) Density Error
| b) Flow up level down
111
Instrument Response during a Casualty
D. DISCUSS how the following factors could cause indication errors on the RPV water level instruments.
3. A rise in drywell temperature
a) Density Error
| b) T up l up
112
Instrument Response during a Casualty
D. DISCUSS how the following factors could cause indication errors on the RPV water level instruments.
4. Reference leg flashing
a) Density Error
| b) Boil up level up
113
Instrument Response during a Casualty
D. DISCUSS how the following factors could cause indication errors on the RPV water level instruments.
5. Flow through the jet pumps
a) Flow Error
| b) Flow up level up
114
Instrument Response during a Casualty
| E. DISCUSS when and why the Floodup and Fuel Zone Range instruments must be compensated for changes in reactor pressure.
1. Floodup < 100# p rx
| 2. Fuel zone corrected when vessel pressurized (calibrated cold)
115
Instrument Response during a Casualty
F. DISCUSS why the Fuel Zone Range instruments can still provide an indication of RPV water level even with the downcomer drained following a large LOCA.
1. As long as jet pump diffusers remain flooded
2. Pressure in jet pump is directly proportional to water level in the core due to manometer effect
3. Boiling in the shroud makes pressure in the jet pump higher
4. High pressure makes jet pump read lower than actual level (conservative)
116
Instrument Response during a Casualty
G. DISCUSS how the RPV water level instruments can continue to indicate some water level even though actual water level is below the variable leg tap.
1. Minimum level based on drywell temperature
117
****IMPORTANT****
Instrument Response during a Casualty
H. STATE when the operator should expect the reference legs on the Wide Range instruments to flash during an emergency depressurization
****IMPORTANT****
1. Wide range ref legs are already heated
2. Somewhere between 500# and 200#
3. When drywell temperature is higher than saturation temperature in RPV
118
Instrument Response during a Casualty
| I. DESCRIBE the response of RPV water level instruments during a LOCA.
1. Regular LOCAs with LP ECCS or a HP ECCS operating, RPV level will remain in Wide Range
2. Big LOCA - .5ft2 of recirc piping
a) Downcomer will drain and level will not be restored even with all ECCS
b) Floodup, narrow and wide range fail low
c) Rely on fuel zone due to jet pump intact manometer, but have errors:
(1) Variable leg temp greater than calibration temp
(2) Two phase flow inside shroud
(3) Ref leg temp greater than calibration temp
(4) Ref leg flashing
(5) Jet pump flow
119
Instrument Response during a Casualty
J. DISCUSS the response of RPV water level instruments to the following conditions:
1. Loss of power supply
a) Fail downscale
120
Instrument Response during a Casualty
J. DISCUSS the response of RPV water level instruments to the following conditions:
2. Rise in ambient temperatures around transmitter
a) Barton: 200F
b) Rosemount: 200F
c) Bailey: 185F
d) Operating characteristics of transistors change with temperature, they have max temperatures to be considered accurate
e) Transmitter and Transistors are outside drywell, not exposed to high T
121
Instrument Response during a Casualty
J. DISCUSS the response of RPV water level instruments to the following conditions:
3. High moisture conditions
a) Corrosion
| b) Not predictable
122
Instrument Response during a Casualty
J. DISCUSS the response of RPV water level instruments to the following conditions:
4. High radiation levels
a) Changes material properties of electronic components
| b) Not predictable
123
Instrument Response during a Casualty
K. DISCUSS how the following systems can be used to obtain an estimation of RPV water level. (LT 20027.011)
1. SRMs
a) Crude level indication based on shielding due to water
b) As it becomes uncovered, reduction of detector output due to lower thermal neutron flux in the void areas (fast neutron flux goes way up)
c) 36”/min with a stopwatch to get height
124
Instrument Response during a Casualty
K. DISCUSS how the following systems can be used to obtain an estimation of RPV water level. (LT 20027.011)
2. IRMs
a) IRMs are less sensitive than the SRMs
| b) Gammas go up when uncovered, IRM power goes up when uncovered
125
****IMPORTANT****
Instrument Response during a Casualty
K. DISCUSS how the following systems can be used to obtain an estimation of RPV water level. (LT 20027.011)
3. LPRMs
****IMPORTANT****
a) LPRMs are useless
126
Instrument Response during a Casualty
K. DISCUSS how the following systems can be used to obtain an estimation of RPV water level. (LT 20027.011)
4. TIPs
a) Isolate on Group 2, and must use a more sensitive detector (ugh)
b) Output goes up when uncovered, barely
127
Instrument Response during a Casualty
L. DISCUSS the use of the following systems during post accident conditions.
1. RPV Pressure Instrumentation
a) Steam tables to get temperature
| b) Magnitude of leak
128
Instrument Response during a Casualty
L. DISCUSS the use of the following systems during post accident conditions.
2. RPV Temperature Instrumentation
a) Can use recirc temps for RPV temp if in service
129
Instrument Response during a Casualty
L. DISCUSS the use of the following systems during post accident conditions.
3. Jet Pump, Core Flow and Core Plate dp Instrumentation
a) LPCI injection can be used for core flow
b) Difference in recirc loop flows
c) Use on recovery
130
Instrument Response during a Casualty
L. DISCUSS the use of the following systems during post accident conditions.
4. Primary Containment Pressure Instrumentation
a) Feel for severity of leak
| b) Rapid reduction could mean primary containment failure
131
Instrument Response during a Casualty
L. DISCUSS the use of the following systems during post accident conditions.
5. Primary Containment Temperature Instrumentation
a) Broader perspective on severity of leak
| b) Water level accuracy when used with instruments
132
Instrument Response during a Casualty
L. DISCUSS the use of the following systems during post accident conditions.
6. Suppression Pool Level Instrumentation
a) Effectiveness of cooling water to RPV as heat sink and supply to ECCS
133
Instrument Response during a Casualty
L. DISCUSS the use of the following systems during post accident conditions.
7. Process Computer
a) Reconstruction of events leading to and following an incident
134
Instrument Response during a Casualty
L. DISCUSS the use of the following systems during post accident conditions.
8. Safety Parameter Display System (SPDS)
a) Multitude of info and centralized data
b) Convenient for not going to back panels or around the plant
c) Post-accident conditions, and allows event to be replayed in simulator for more information gathering not done in situ for cause and consequences
135
```
Transient Analysis
A. Given chart recorded traces indicating the plant response to various transients, determine the following:
1. Initiating Cause of Event
2. Cause of SCRAM (if it did)
3. Sequence of Major Events
```
1. Initiating cause of event
2. Cause of SCRAM (if it did)
a) Black APRM line goes straight down on a SCRAM
b) Check if SCRAM happened later (not as initiator)
3. Sequence of major events
a) POWER - Derate only on Recirc pump trip
(1) One pump has % core flow sharp drop and recovery
(2) Two pumps
b) PRESSURE
(1) Loss of Feed has slow slope pressure drop
(2) Turb trip w/o BPV has perfect sawtooth pressure
c) RWL
(1) Turb Trip has HUGE RWL spike
(2) MSIV closure has upside down M RWL similar to turb trip
(3) EHC has late RWL spike
d) FLOW
(1) Recirc Cntrlr fail has high % core flow endgame
(2) FW Cntrlr fail has highest FW flow w/o immediate core flow drop
e) IF not anything yet, must be a plain SCRAM
(1) Power goes down first, nothing else interesting
136
Plant Safety – Accident Analysis
| A. STATE the objective of the Plant Safety Analysis.
1. To evaluate the ability of the plant to operate without undue hazard to the health and safety of the public
137
Plant Safety – Accident Analysis
| B. STATE the three basic groups of events which are evaluated by the Plant Safety Analysis.
1. Anticipated Operational Occurrences (AOOs)
2. Accidents
3. Special Events
138
Plant Safety – Accident Analysis
| C. DISCUSS unacceptable results for Anticipated Operational Occurrences evaluated by the Plant Safety Analysis.
1. Release to the environment
2. Fuel cladding failure
3. Pressure boundary stress exceeding allowable stress
4. Exceeding suppression pool heat capacity temperature limit
139
Plant Safety – Accident Analysis
| D. DISCUSS unacceptable results for accidents evaluated by the Plant Safety Analysis.
1. Release to the environment
2. Catastrophic failure of fuel cladding, including fragmentation of cladding and excess enthalpy
3. Pressure boundary stresses exceeding those allowed
4. Containment stresses exceeding those allowed
5. Overexposure to radiation of operators in MCR
140
Plant Safety – Accident Analysis
| E. DEFINE the term "Anticipated Operational Occurrence" as stated in the FSAR.
1. Single errors or failures that can be reasonably expected during any normal or planned mode of plant operations
141
Plant Safety – Accident Analysis
| F. DEFINE the term "operator error" as stated in the FSAR.
1. Deviation from procedures as event initiator (set of actions included)
142
Plant Safety – Accident Analysis
| G. DEFINE the term "accident" as stated in the FSAR
1. Postulated events that may affect one or more of the radioactive material barriers which are not expected during the life of the plant
143
Plant Safety – Accident Analysis
| H. DEFINE term "design basis accident (DBA)" as stated in the FSAR.
1. Greater than any other accident
144
Plant Safety – Accident Analysis
| I. STATE the four design basis accidents which are evaluated by the Plant Safety Analysis.
1. CRDA
2. LOCA
3. MS Line Break Accident
4. Refueling Accident
145
Plant Safety – Accident Analysis
J. DISCUSS the plant safety analysis for the following DBAs, including possible causes, assumptions, general course of events during the accident, and expected results:
1. Control Rod Drop Accident
a) High worth control rod fully out of the core
b) From the top to the bottom
c) APRM 120% signal SCRAMs in less than 5 sec
d) <280cal/g (425cal/g fuel vaporizes)
146
Plant Safety – Accident Analysis
J. DISCUSS the plant safety analysis for the following DBAs, including possible causes, assumptions, general course of events during the accident, and expected results:
2. Loss of Coolant Accident (LOCA)
a) LOSP and suction recirc line break simultaneously (SCRAM and Gp1)
b) ECCS and LOCA timers start systems, ADS, CS
c) Core rapidly refloods and heatup will be terminated within 100sec
d) <2200F, no pins fail, DW pressure peaks at 46.7# in 10sec, then goes down
147
Plant Safety – Accident Analysis
J. DISCUSS the plant safety analysis for the following DBAs, including possible causes, assumptions, general course of events during the accident, and expected results:
3. Main Steam Line Break Accident
a) 5.5sec closure of full break
b) SCRAM on 90% open MSIV
c) <2200F, no fuel damage, 52300lbm lost
148
Plant Safety – Accident Analysis
J. DISCUSS the plant safety analysis for the following DBAs, including possible causes, assumptions, general course of events during the accident, and expected results:
4. Refueling Accident
a) Failure of lifting mechanism and dropped fuel on core
b) High power for 1095 days
c) 24 hours to S/D, C/D, RPV head removed, upper internals removed
d) Dropped fuel assembly from 30ft at 40ft/sec
e) Hits 4 assemblies on first bounce, 24 on second bounce, plus 3rd bounce
f) Worst DBA
g) 14800 Curies of Iodine and Noble gases released
149
Plant Safety – Accident Analysis
K. DESCRIBE the methods for evaluating damage to the following radioactive material barriers:
1. Fuel Cladding
a) 99.9% not get boiling by MCPR
| b) 2200F by MAPRAT
150
Plant Safety – Accident Analysis
K. DESCRIBE the methods for evaluating damage to the following radioactive material barriers:
2. Reactor Coolant Pressure Boundary
a) 1375# peak internal pressure
| b) 1325# Steam dome safety limit
151
Plant Safety – Accident Analysis
K. DESCRIBE the methods for evaluating damage to the following radioactive material barriers:
3. Containment
a) Remain below max values specified by mechanical design
152
Plant Safety – Accident Analysis
| L. STATE the reason for special design considerations associated with High Energy Line Breaks (HELB).
1. FSAR Supplement 15A
a) HELB outside primary containment
b) Bring and maintain reactor in a safe shutdown condition
c) Radioactive releases will not exceed 10CFR100 values
