57
HTGR Technology Course for the Nuclear R l t C i i Regulatory Commission May 24 – 27 2010 May 24 27, 2010 Module 14 HTGR Accident Analyses Fred Silady Technology Insights Slide 1 1
HTGR Technology Course for the Nuclear R l t C i i Regulatory Commission
May 24 – 27 2010May 24 27, 2010
Module 14HTGR Accident Analyses
Fred SiladyTechnology Insights
Slide 11
Outline
• Licensing basis event selection• Licensing basis event selection
• Event types and accident analysis resultsEvent types and accident analysis results– Challenges to core heat removal– Challenges to control of heat generationg g– Challenges to control of chemical attack
Slide 22
Context of Licensing Basis Events within Elements of Industry-Proposed Licensing Approach
• What must be met– Top Level Regulatory Criteria (TLRC)
• When TLRC must be met– Licensing Basis Events
• How TLRC must be met– Safety Functionsy– SSC Safety Classification– Regulatory Design Criteria
• How well TLRC must be met– Deterministic DBAs
Defense in Depth
Slide 33
– Defense-in-Depth– Regulatory Special Treatment
Industry Proposed Process for LBE Selection (1/3)
1. Define region boundaries
2 C i k t lt t i d 2. Compare risk assessment results to region dose limits
3 Identify as AOOs families of events in AOO 3. Identify as AOOs families of events in AOO region that could exceed 10CFR20 offsite doses if certain equipment or design features had not been selected
4. Evaluate AOO consequences including t i ti d th t uncertainties and assure that mean
consequences meet 10CFR20 offsite dose limits
Slide 44
Industry Proposed Licensing Basis Event Regions
1.E+00
1.E+01
YEA
R)
ANTICIPATED OPERATIONALOCCURRENCE(A00) REGION
10CFR20(Measured in TEDE at Controlled
Area Boundary)
U t bl
1.E-02
1.E-01
Y (P
ER P
LAN
T Y
10% OF 10CFR50.34(Measured in TEDE at EAB)
DESIGN BASIS EVENT
(MEAN FREQUENCY 1.0E-02)
Unacceptable
1.E-04
1.E-03
AN
FR
EQU
ENC
Y
(DBE) REGION
10CFR50.34(Measured in TEDE at EAB)
(MEAN FREQUENCY 1.0E-04)
Acceptable
1.E-06
1.E-05
SEQ
UEN
CE
MEA
(MEAN FREQUENCY 5.0E-07)
BEYOND DESIGN BASIS EVENT (BDBE)
REGION
PROMPT MORTALITYSAFETY GOAL
(Measured in Whole Body Gamma Dose in the Region to
1.6Km)DESIGN GOAL PLUME PAG
(M d i TEDE)
1 E-08
1.E-07
EVE
NT
S (Measured in TEDE)
Note: The Safety Goal limit is plotted at the EAB for illustration purposes; otherwise it would be off the chart.
Slide 55
1.E-081.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1.E+01 1.E+02 1.E+03 1.E+04
DOSE (REM) AT EXCLUSION AREA BOUNDARY
Industry Proposed Process for LBE Selection (2/3)
5. Identify as DBEs families of events in DBE region that could exceed 10CFR50.34 doses if certain equipment or design features had not been selected
6. Evaluate consequences of any DBEs with upper bound uncertainty in the AOO region and assure that the mean consequence meets 10CFR20 offsite dose limitsconsequence meets 10CFR20 offsite dose limits
7. Evaluate DBE consequences including uncertainties and assure that the mean consequence of each meets the EPA Protective Action Guidelines at the EAB site boundary (design goal)Action Guidelines at the EAB site boundary (design goal)
8. Select deterministic Design Basis Accidents (DBAs) from the DBEs by assuming that only SSCs relied on to meet 10CFR50.34 (th l ifi d S f t R l t d) il bl(those classified as Safety Related) are available
9. Evaluate DBEs and deterministic DBA consequences including uncertainties and assure that the upper bound consequences
Slide 66
pp qmeet 10CFR50.34 offsite dose limits
Pebble Bed Example of Safety Classification for Core Heat Removal Function
Are SSCs Available and Sufficient to Remove Core Heat in the DBE?
Alternative Sets of SSCs
DBE 1c DBE 2b DBE 6c DBE 7a DBE 7b DBE 11b
SSCs Classified as Safety Related?
Reactor PCU ACS
No No No No No No
Reactor No No No Yes No NoSBS ACS
No No No Yes No No
Reactor CCS ACS
No Yes No Yes Yes Yes
Reactor Reactor vessel Active RCCS
ACS
Yes Yes Yes Yes Yes Yes
ReactorReactor Reactor vessel Passive RCCS
Yes Yes Yes Yes Yes Yes Yes
Reactor Reactor vessel
Building & ground Yes Yes Yes Yes Yes Yes
Slide 77
g g
Note: Italics indicates response during DBE
Industry Proposed Process for LBE Selection (3/3)
10. Identify as BDBEs the dose-dominant families of events in BDBE region
11. Evaluate consequences of any BDBEs with upper bound uncertainty in the DBE region and assure that the upper bound consequence of each meets 10CFR50.34 offsite dose limits
12. Evaluate BDBE consequences including uncertainties and assure that the mean consequence of each meets the EPA assure that the mean consequence of each meets the EPA Protective Action Guidelines (design goal)
13. Evaluate overall cumulative risk including all LBEs and assure NRC safety goal quantitative health objectives (51FR130) are NRC safety goal quantitative health objectives (51FR130) are met
14. Assure that residual risk is negligible
Slide 88
Prismatic MHTGR Licensing Basis Events
Slide 99
Outline
• Licensing basis event selection
• Event types and accident analysis results– Challenges to core heat removal– Challenges to control of heat generation– Challenges to control of chemical attack
Slide 1010
Modular HTGR Accident Safety Evaluations
• Challenges to core heat removal– Loss of heat transport (HTS) & shutdown forced cooling
systems (SCS/CCS)systems (SCS/CCS)(Pressurized conduction cooldown or PLOFC)
– Depressurization and Loss of HTS & SCS/CCS(D i d d ti ld DLOFC)(Depressurized conduction cooldown or DLOFC)
• Challenges to control heat generation– Accidental control rod withdrawal– Station blackout without trip
• Challenges to control chemical attack– Water/steam ingress from SG tube break– Air mixture ingress from RB following HPB leaks/breaks
Slide 1111
Air mixture ingress from RB following HPB leaks/breaks
Functions for Control of Radionuclide Release
Maintain Control of Radionuclide Release
Control Control PersonnelControl Radiation
Control Personnel Access
Control Radiation from Processes
Control Radiation from Storage
Control Radiation from Core from Processes Storagefrom Core
Control Radiation Transport
Control Direct Radiation
Control Transport from Site
Control Transport from Reactor Building
Control Transport from HPB
Control Transport from Core
Retain Radionuclides in Fuel Elements
Control Radionuclides in Fuel Particles
Denotes Minimum Functions to Meet
10CFR50 34Remove Core Heat Control Core Heat Control Chemical
Slide 1212
10CFR50.34Remove Core Heat Control Core Heat Generation
Control Chemical Attack
Pebble Bed Relative I-131 Inventories within HPB
SourceI-131
400MWt InventorySource 400MWt Inventory (Ci)
Circulating activity <<1g yPlateout on internal Helium Pressure Boundary (HPB) surfaces
<1
Uranium contaminated fuel particles ~20Uranium contaminated fuel particles ~20Failed and defective fuel particles ~580Intact fuel particles 1 x 107
Slide 1313
Circulating Activity, Plateout, and Dust Release
• Circulating activity– Released from HPB with helium in minutes to days as a result of HPB
leak/break – Amount of release depends on location and any operator actions to
isolate and/or intentionally depressurize
Liftoff of plateout and resuspension of dust• Liftoff of plateout and resuspension of dust– Liftoff physical and chemical phenomena include:
• Particulate entrainment: removal of dust, oxidic and metallic particles from surfaces
• Desorption: removal of atoms or molecules sorbed from surfaces• Diffusion: transport of fission or activation products from surface inward or
to and from particulates• Aerosol formation: mechanism by which the particulates are formedAerosol formation: mechanism by which the particulates are formed
– For large breaks partial release from HPB with helium relatively quickly (minutes)
– Amount of release depends on HPB break size that results in surface h f t th l ti fl
Slide 1414
shear forces greater than normal operation flows
Pebble Bed Main Power System (MPS) Pressure Following HPB Leaks and Breaks (400MWt)
10000
7000
8000
9000 230mm
65mm
10mm
5000
6000
ssur
e (k
Pa)
Moderate break
2000
3000
4000
Pre
Small leak
Small break
0
1000
0.000 0.001 0.010 0.100 1.000 10.000 100.000
Small leak
Slide 1515
Time (hr)
Pebble Bed Shear Force Ratio (SFR) Results for Range of HPB Leak/Break Sizes at Core Inlet Plenum (CIP)
SFR vs. CIP Equivalent Break Size for 500MWt PBMR Design
10mm 30mm 100mm 230mm
Reactor Inlet 0.03 0.04 0.83 2.3
Reactor Outlet 0.02 0.02 0.99 1.0Reactor Lower Volume 0.07 0.08 0.99 2.4
CCS I l t C ti 0 02 0 02 0 95 1 0CCS Inlet Connection 0.02 0.02 0.95 1.0
IHX Inlet 0.01 0.01 0.23 1.0Circulator Outlet 0.01 0.01 0.02 1.0
Breaks ≤100mm have SFR <1: insignificant dust resuspension and liftoff
Slide 1616
Removal of Core Heat Accomplishedby Passive Safety Features
• Small thermal rating/low power density– Limits amount of afterheat– Low linear heat rate
• Core annular/cylindrical geometryH t l b i d ti d – Heat removal by passive conduction and radiation mechanisms
– High heat capacity graphiteHigh heat capacity graphite– High temperature core materials
Slide 1717
SSCs Supporting Core Heat Removal
Active and passive engineered systems– Heat Transport System (HTS)/ Main Power
System (MPS)Shutdown Cooling System (SCS)/ Core – Shutdown Cooling System (SCS)/ Core Conditioning System (CCS)
– Helium Purification System Post Accident Train y(Pebble Bed HPS PAT)
– Reactor Cavity Cooling System (RCCS)A ti d• Active mode
• Passive mode
Slide 1818
Passive Heat Transfer to Air-Cooled RCCS
Slide 1919
Prismatic DCC Peak Fuel Temperatures
Slide 2020
Heat Transfer in the Pebble Bed
Slide 2121
Pebble Bed Fuel Temperatures with Forced Core Cooling (CCS) & Passive Conduction Cooldown (400MWt)
Maximum Pebble Fuel Temperature
1800
1400
1600
e [°
C]
DLOFC, 100 kPa
PLOFC, 6000 kPa
CCS, 100 kPa
CCS 1000
800
1000
1200
empe
ratu
re CCS, 1000 kPa
400
600
800
Max
imum
t
0
200
0.00 10.00 20.00 30.00 40.00 50.00 60.00 70.00 80.00 90.00 100.00
Slide 2222
Time [hrs]
Pebble Bed Temperatures for PLOFC (400MWt)
Decay Heat Removal at 6000 kPa using Passive Means
1400
1000
1200
e [°
C]
Max Fuel
Max CB
Max RPV
600
800
1000
Tem
pera
ture
CB Transient Limit
400
600
Max
imum
T
RPV Transient Limit
0
200
0.00 10.00 20.00 30.00 40.00 50.00 60.00 70.00 80.00 90.00 100.00
Ti [h ]
Slide 2323
Time [hrs]
Pebble Bed DLOFC Core Average Fuel Temperature (500MWt)
1100
1150
1000
1050
1100
C)
100 mm (Tdp = 0.15hr)10 mm (Tdp = 14.4hr)5 mm (Tdp = 58hr)4 mm (Tdp = 90.5hr)3mm (Tdp = 156hr)
900
950
el te
mpe
ratu
re (C
800
850Fue
700
750
0 15 30 45 60 75 90 105 120 135 150 165 180 195 210 225 240 255 270 285 300
time (h)
Slide 2424
Pebble Bed DLOFC Maximum Fuel Temperature (500MWt )
1600
1650
1700
1400
1450
1500
1550
600
C)
100 mm (Tdp = 0.15hr)10 mm (Tdp = 14.4hr)5 mm (Tdp = 58hr)4 mm (Tdp = 90.5hr)
1250
1300
1350
1400
el te
mpe
ratu
re (C
( )3mm (Tdp = 156hr)
1050
1100
1150
1200Fue
900
950
1000
0 15 30 45 60 75 90 105 120 135 150 165 180 195 210 225 240 255 270 285 300
Slide 2525
time (h)
Pebble Bed Spatial DLOFC Maximum Fuel Temperature (53hr) for 100mm Break (500MWt)
177
185
143
152
161
169
axis
(cm
)
118
125
133
r
22 71 121 145 170 219 268 318 367 416 465 515 564 613 662 712 761 810 859 909 958 1007 1045 1083 1121 1158109
z axis (cm)
500-600 600-700 700-800 800-900 900-1000 1000-1100 1100-1200 1200-1300 1300-1400 1400-1500 1500-1600 1600-1700
Slide 2626
Pebble Bed DLOFC Temperatures Showing % of Fuel Volume at 50 Hr (500MWt)
10
11
12
7
8
9
in in
terv
al
4
5
6
% o
f cor
e vo
lum
e i
1
2
3
0350 450 550 650 750 850 950 1050 1150 1250 1350 1450 1550 1650
temperature interval (C)
3 mm 5 mm 10 mm 100 mm
Slide 2727
Delayed Fuel Release Mechanisms
• Partial release from contamination, initially failed, or defective particles when temperatures exceed normal operation levels and from particles that fail during the event
• Timing of release is tens of hours to days• Inventory is much larger than circulating activity and liftoff• Amount of release from fuel depends on fraction of core p
above normal operation temperatures for given times and on radionuclide volatility– Governed by amount of forced cooling
D d t h th ll l k l b k– Dependent on whether small leak or large break• Amount of release from HPB depends on location and size of
leak/break and on timing relative to expansion/contraction of gas mixture within the HPBgas mixture within the HPB– Small leaks have greater releases from HPB– Releases cease when the HPB internal system temperature
decreases due to core temperature cooldown
Slide 2828
Prismatic Cumulative RN Releases from Fuel During DCC (350MWt)
Slide 2929
Prismatic Cumulative RN Releases from HPB During Small Leak DCC (350MWt)
Slide 3030
Prismatic Cumulative RN Releases from RB During Small Leak DCC (350MWt)
Slide 3131
Pebble Bed DLOFC Dose as a Function of HPB Leak/Break Size for Vented RB (500MWt)
1000
10000
10 CFR 50.34 Limit = 25 rem TEDE
EPA PAG Limit = 1 rem TEDE
100
1000
E [m
rem
]
Direction of increasing effect of
BDBEDBE
10
dary
Dos
e TE
D
gblowdown driving
out the delayed fuel release
1
Site
Bou
nd
From 3mm to 2mm blowdown time is
greater than time of
0.01
0.1
Direction of increasing Shear Forces during Blowdown to lift-off plateout and dust
greater than time of fuel release allowing more decay before
release
Slide 3232
1 10 100 1000
Break Diameter [mm]
Outline
• Licensing basis event selection
• Event types and accident analysis results– Challenges to core heat removal– Challenges to control of heat generation– Challenges to control of chemical attack
Slide 3333
HTGR Control of Heat Generation
• Continued functioning of reactor shutdown system only necessary for long-term shutdown– Negative temperature coefficient for reactivity
• Temperature differential of 750K maintained between operational and maximum allowable fuel temperature
• Reactor shuts itself down before maximum fuel temperature reached
– Limited excess reactivity
– Integrity of core structures• Ceramic core structures and fuel elements• Simple and robust core structure design
Slide 3434
HTGR Reactivity Insertion Mechanisms
• Range of initial conditions of core temperature, core reactivity, control rod insertion, Xenon decay timesdecay times
• Control rod and control rod group withdrawalControl rod and control rod group withdrawal
• Removal of RSS small absorber spheresp
• Increased moderation from water ingress
• Core compaction from seismic events (pebble bed)
Slide 3535
bed)
AVR Test Demonstrated that Nuclear Reaction Terminates with Loss of Forced Cooling
Slide 3636
MHTGR Analysis Showed Similar Behavior to AVR Test
Slide 3737
Prismatic Accidental Control Rod Withdrawal AnalysisDemonstrates Mitigation of Reactivity Event
• Spurious rod withdrawal initiated from 100% powerpower
• Transient analyzed with two protection system responses– Normal control rod trip
Backup reserve shutdown control material trip – Backup reserve shutdown control material trip (rod trip suppressed)
• Reactor thermal and nuclear characteristics provide inherent limit on power increase rate and magnitude
Slide 3838
magnitude
Prismatic Reactor Temperatures Well Below Limits during Accidental Control Rod Withdrawal
Slide 3939
Prismatic Core Temperatures Maintained at Safe Levels with and without Reactor Trip
Slide 4040
Outline
• Licensing basis event selection process
• Event types and accident analysis results– Challenges to core heat removal– Challenges to control of heat generation– Challenges to control of chemical attack
Slide 4141
Control of Water Chemical Attack
• Non-reacting coolant (helium)
• Water-graphite reaction:– endothermic– requires temperatures exceeding normal
operation (>700°C)slow reaction rate– slow reaction rate
• Graphite and silicon carbide coatings protect fuelp g p
Slide 4242
Prismatic Power During SG Tube Rupture Without Forced Cooling (350MWt)
Slide 4343
Prismatic Pressure During SG Tube Break Without Forced Cooling (350MWt)
Slide 4444
Prismatic Graphite Oxidation During SG Tube Break Without Forced Cooling (350MWt)
Slide 4545
Control of Air Chemical Attack
• Non-reacting, pressurized coolant (helium)
Ai i li it d • Air ingress limited – HPB configured with three Class 1 vessels
– HPB piping diameter limited (~65mm dia)
– HPB leaks/breaks result in venting of most RB air
• Slow oxidation rate of core support and reflector nuclear grade graphitenuclear grade graphite
• Ceramic coated particles embedded within fuel l t
Slide 4646
elements
Conditions Required for Self-Sustained Oxidation of Nuclear-Grade Graphite
• Heat generation from exothermic oxidation must exceed heat loss by conduction, convection, radiationradiation
• Heat generation rates are low because:Heat generation rates are low because:– Very low concentrations of volatiles and catalytic
impuritiesReaction rates limited at higher temperatures by oxygen – Reaction rates limited at higher temperatures by oxygen diffusion across boundary layer and into graphite
• Heat losses are high because:– High thermal conductivity and emissivity– Low-temperature air gas mixture provides convective
Slide 4747
p g pcooling
Progression of Air Ingress Events
• Overall oxidation rate determined by rate of air supply
F i ti tl li it fl t– Friction greatly limits flow rate– Flow rate further limited as core heats up because
viscosity increases with temperature– Eventual core cooling limits oxidation to negligible level– Graphite mass loss is a few percent at most and limited
to lower plenum and reflectors• Radioactivity released by graphite oxidation is
smallRelatively low levels of radioactivity in graphite– Relatively low levels of radioactivity in graphite
– Radiological consequences only marginally greater than conduction cooldown w/o air ingress
Slide 4848
Prismatic Slow Oxidation of Graphite Limited by Air Mass Transfer and Core Temperataures (350MWt)
Assumes 100% air from RB after helium depressurizationafter helium depressurization
Slide 4949
PBMR Reactor Building Vent Pathway Influence on Air Mixture Ingress (500MWt)
NHSB (Reactor B ilding)
Reactor Top Cavity (RTC)
RB Vent
NHSB (Reactor Building) Boundary - Leak Rate 100 vol%/d
R t
Top Head / RCCS / HVAC Area
PRS EngineeredFeatures Area
little or no partici-pation in LBE & BDBE events
(RTC)
ReactorCavity IHX Access
Area
IHX Area
small and medium PHBP breaks - maximum 100 mm
little or no participation in LBE & BDBE events
Red Arrows show Engineered Vent Path consisting of rupture disks
Spent Fuel and AuxiliariesArea
little or no
small, medium andlarge PHBP breaks - maximum 270 mm (LBE)
1000 mm (BDBE)
DEG SHBP breaks(IHX area only)
consisting of rupture disks, dampers, and otherfeatures
FHSS Area
CCS and HPS Area
PRS
little or no participation in LBE & BDBE events
1000 mm (BDBE)
Slide 5050
FHSS Area Vent Path
Pebble Bed Gas Mixture in RB for 100mm HPB Break RB Vent Fails to Reclose Case
Compartment He Mass Fraction
0.8
1.0
actio
n
He in Reactor Top Cavity: 0.94
0 2
0.4
0.6
He
Mas
s Fr
a
0.0
0.2
0 1200 2400 3600
Time (s)
He in Vent Plenum: 0.35
Time (s)
Reactor Inlet Reactor Top Cavity FIP Compartment
0 9 mass fraction He (i e 10% air content by mass) in RTC after first hour
Slide 5151
~0.9 mass fraction He (i.e.,10% air content by mass) in RTC after first hour
Pebble Bed Gas Mixture Ingress for 100mm HPB Break RB Vent Fails to Reclose Case
NHSS and Reactor Air Mass Fraction
0.025
0 015
0.020
0.025
ctio
n Mass fraction air in reactor: 0.022
0 005
0.010
0.015
Air
Mas
s Fr
a
0.000
0.005
0 50 100 150 200 250 300
Time (hr)
Assumed failure to close RB damper
Reactor Inlet Reactor Outlet Reactor Lower Volume CCS Inlet Connection Void Above Core Below Outlet Slots
Slide 5252
~2% air content by mass in Reactor after 300hrs
Role of Reactor Building in Safety Design
• Required safety function of RB is to structurally protect HPB, Reactor, and RCCS from external
t d h devents and hazards• RB provides additional radionuclide retention and
limits air available for ingress after HPB limits air available for ingress after HPB depressurization
• Vented design superior to pressure retaining d i f G h t i tidesign for HTGR characteristics– Less air available in gas mixture for ingress to reactor
after helium depressurization and venting– Pressurized non-condensable helium not available to
transport RNs from delayed fuel release by leakage or subsequent RB failure
Slide 5353
Comparison of RB Alternatives to PAG Sheltering Dose at EAB
100,000
10,000
d D
ose
[mre
m]
EPA PAG Limit = 5 rem Thyroid
1,000
Bou
ndar
y Th
yroi
d
100Site
B
10No RB Alternative
1a/1bAlternative 2 Alternative 3a Alternative 2 or
3a with failure ofadded features
Alternative 4a Alternative 4awith late CF
Vented Alt 1
Pressure Retaining
Pressure Retaining with late fail re
Vented Alt 2
Vented Alt 3a
Slide 5454
failure
Important HTGR Safety Paradigm Shifts• The fuel, helium coolant, and graphite moderator are chemically
compatible under all conditions
• The fuel has very large temperature margins in normal operation and • The fuel has very large temperature margins in normal operation and during accident conditions
• Safety is not dependent on the presence of the helium coolant
• Response times of the reactor are very long (days as opposed to seconds or minutes)
• There is no inherent mechanism for runaway reactivity excursions or power excursions
• The HTGR has multiple, nested, and independent radionuclide barriersThe HTGR has multiple, nested, and independent radionuclide barriers
• An LWR-type containment is neither advantageous nor necessarily conservative.
Slide 5555
Summary
• HTGR LBEs selected systematically using risk insights
• Modular HTGR safety design focuses on radionuclide retention at the source within the fuel
• Challenges to the radionuclide retention grouped by the three key functions that are met with the inherent
h t i ti f th f l l t d d t d th characteristics of the fuel, coolant, and moderator and the passive reactor configuration
• Modular HTGR accident time scales are long and the phenomena are amenable to mechanistic evaluations
Slide 5656
Suggested Reading• NGNP Licensing Basis Event Selection White Paper (~June
2010).
• “Preliminary Safety Information Document for the Standard MHTGR,” DOE-HTGR-86024, Rev. 13, September 1992, ML093560560.ML093560560.
• “PBMR Reactor Building Functional and Technical Requirements and Evaluation of Reactor Embedment ” NGNPRequirements and Evaluation of Reactor Embedment, NGNP-NHS 100-RXBLDG, Rev 0, Westinghouse Electric Company LLC, September 2008.
• “PBMR Plant Level Assessments Leading to Fission Product Retention Allocations,” NGNP-FPA-RPT-001, Rev 0, Westinghouse Electric Company LLC July 2009
Slide 5757
Westinghouse Electric Company LLC, July 2009.
