plan b
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“Plan-B”An Alternative Liquidation* Strategy
of Fukushima Daiichi NPP
May 21, 2011Satoshi Sato
Satoshi.sato@iacdc.com
International Access Corporation
*: A term “liquidation” is used in this document to generally mean various activities directly and indirectly associated with restoration of safe state of each affected reactor in Fukushima Daiichi NPP. This follows a precedent in which workers involved in the emergency actions on the Chernobyl site during the accident and the subsequent clean-up operations were called “Liquidators”.
2
BWROG BWR Owners Group
CCI Core-Concrete Interaction
EPG Emergency Procedure Guidelines
FHM Fuel Handling Machine
FP Fission Product
GTCC Greater Than Class C (Cask for High Level Radiation Waste)
ISFSI Independent Spent Fuel Storage Installation
NPP Nuclear Power Plant
OHC Overhead Crane
RPV Reactor Pressure Vessel
SFP Spent Fuel Pool
SGTS Standby Gas Treatment System
SNF Spent Nuclear Fuel
Abbreviations
3
4
5
Contents
• Present Status• “Plan-A” Ever discussed?
– Difficulties to dismantle BWR reactor with damaged core
– Difficulties to retrieve spent fuel from degraded Reactor Bldg.
– Conventional Decommissioning Processes– Definition of “Plan-A”
• Developing “Plan-B”– Worst Case Scenario
6
Contents (cont’d)
• “Plan-B”– Liquidation Strategies for Fukushima NPP Reacto
rs and SFPs– Water Treatment and Entombment– On-Site Above-Ground Repository– Design Beyond Millennium– Beyond “Liquidation
• Next Step
7
Present Status
• Reactor Status– Reactor Core
– Reactor Pressure Vessel
– Primary Containment
– Reactor Building (Secondary Containment)
– Residual Heat Generation
• Spent Fuel Pool Status– Fuel Integrity
– Pool
– Residual Heat Generation
8
Summary - Reactor
UnitReactor
CoreReactor
Pressure VesselPrimary
ContainmentReactor Building
1 Totally destroyed
Barrier integrity no longer maintained. Bottom Head Penetrations severely damaged.
Barrier integrity no longer maintained. Details not confirmed.
Original Function as Secondary Containment totally lost due to H2 explosion on Refueling Floor. Remaining part of building still reasonably good.
Overhead Crane (OHC) and Fuel Handling Machine (FHM) not available.
2 Ditto Ditto
Barrier integrity severely degraded due to H2 explosion inside or outside Torus.
Function as Secondary Containment still reasonably maintained even after H2 explosion. OHC and FHM still fully functioning.
3 Ditto Ditto Same as Unit 1Same as Unit 1, except that some portions lower than Refueling Floor also degraded due to H2 explosion
4 Empty Not affected Not affected Ditto
Function Barrier Integrity
Severely damaged Severely damaged
Severely damaged Possibly still partly maintained but not confirmed
9
Unit 4 Unit 3 Unit 2Unit 1
State of Reactor Building, Unit 1 to 4 looking from east as of March 20
10
Unit 3 Unit 4
State of Reactor Building, Unit 3 and 4 looking from west as of March 20
11
Reactor Core, Pressure Vessel, and Primary Containment
• Current Degree of Degradation of each FP Barrier– Melt-Down through Core Plate: No Doubt
– Leakage of RPV Bottom Head: No Doubt
– Gross Failure of RPV Bottom Head: Not Very Likely
– Leakage of Primary Containment: Highly Likely
– Gross Failure of Primary Containment:
• Unit 1 and Unit 3: Not Very Likely
• Unit 2: Highly Likely (Suppression Chamber)
– Major Core-Concrete Interaction ( CCI ) : Not Very Likely
– Melt-Down through Man-made Rock (Basemat): Not Likely
12
12~
13m
~33.5m
ID 8.9m
~46m
~23.5m
~15m
~11m
~40m
16~
17m
Refueling Floor
Rx. Bldg. (Secondary Containment)
Primary Containment
Reactor Pressure Vessel
Suppression Chamber (part of Primary Containment)
Drywell
Pedestal
Typical Configuration ( Unit 3, 4)
13
Melt-Down through Core Plate
Predicted to occur 2 hours following complete loss of cooling capability.
Several previous experiments suggested steam explosion not likely.
Core Shroud
Core Plate
Reactor Core
Molten Core
Water
No Doubt
14
Further Melt-Down through Core Plate
Unit Duration
1 14h09m
2 06h29m
3 06h43m
Actual Complete Loss of Cooling Capability
(Official Announcement by Government on May 16, 2011)
No Doubt
15
Degradation of Reactor Pressure Vessel Bottom Head
Creep rupture begins to occur at ~240-deg C below melting point (1500-deg C) of vessel material (low alloy steel), allowing some leakage of highly contaminated water containing fractured pieces of fuel pellets.
Highly Likely
16
Locations of Potential Leakage (Typ.)
Vulnerability of Bottom Head Leakage
17
Further Degradation of Reactor Pressure Vessel Bottom Head
Drywell Sump Pit
Pedestal
Pedestal DoorwayPossible
18
Major Degradation of Reactor Pressure Vessel Bottom Head and Core-Concrete Interaction (CCI), Resulting in Significant Amount of Release of Radioactive Aerosol
Pedestal Doorway
Pedestal
H2O,
CO2
H2O,
CO2
H2, CO
Aerosol
AerosolAerosol
Aerosol
Not very likely, but could have happened depending on cooling evolution during early stage.
19
Beginning of Primary Containment Failure
Pedestal DoorwayPedestal
Aerosol
Aerosol
Aerosol
Aerosol
H2, CO
H2O,
CO2
H2O,
CO2
Not very likely, but could have happened depending on cooling evolution during early stage.
20
Pedestal Doorway
Pedestal Wall
Source: NUREG/CR-6042 Rev.2
21
Beginning of Primary Containment Failure
Aeros
ol
Not very likely, but could have happened depending on cooling evolution during early stage.
22
Not likely
Failure due to Creep Rupture
Gross Failure of Primary Containment due to Steam Explosion
23
Gross Failure of Primary Containment due to Melt-Down
Aeros
ol
Aeros
ol
Aeros
ol
Aeros
ol
Not likely
24
Complete Melt-Down through Man-made Rock (Basemat)
Not likely
Man-made Rock
25
Unit 1 2 3 4
# of Fuel Assembly in Rx. 400 548 548 0
Electrical Output (MWe) 460 784 784 0
Thermal Output (MWt) 1,380 2,381 2,381 0
Estimated Residual Heat (MWt)0.1% of rated Thermal Output
1.4 2.4 2.4 0
Residual Heat Generation
2 months after
shutdown
26
Summary - SFP
• Fuel Integrity– No conclusive information so far.– Potential thermal damage. (Units 3 and 4)– Potential mechanical damage. (Units 1 to 4)
• Pool– Details unknown, but apparently no major damage. – Potential thermal damage due to overheating.
(Units 3 and 4)– Potential mechanical damage due to earthquake and/or H2
explosion.
(Units 1 to 4)
27
Unit 4 SFP
Top view of fuel rack by remote underwater TV camera. Difficult to draw any conclusion about fuel integrity only based on this information. Fuel inspection by “sipping” is warranted.
28
Unit 1 2 3 4
Number of Fuel Assembly
New Fuel Storage Vault 100 28 52 204
Spent Fuel Pool 292 587 514 1331
Hottest Spent Fuel Discharged
(Date of beginning of last refueling outage)3/25/’10 9/16/’10 6/19/’10 11/30/’10
Estimated Residual Heat Generation Rate (MWt) 0.07 0.46 0.23 1.8
Residual Heat Generation
29
Intentionally left blank
30
“Plan-A”Has it ever been discussed?
• Efforts to achieve so-called “Cold Shutdown” have been being made as the single topmost priority.– Originally, this term appears to specifically mean a state par
tially submerging the reactor core from both inside and outside Reactor Pressure Vessel (RPV) to keep the entire metal surface temperature below 200-deg F as defined by Tech Spec.
– However, after realizing that the barrier integrity of the RPV has been excessively challenged, this term now means a submersion only from outside RPV.
– In more common wording, “Drywell Flooding”. – Consistent with the intent per BWROG’s EPG.
31
Drywell Flooding
32
Issues – Public Perspectives
• “Cold Shutdown” is apparently considered to be only the first milestone of the entire “Liquidation” program . It is not a solution or a goal. Nothing has been told to the public beyond that point. Both tax-payers and rate-payers are concerned about the cost and schedule of the entire program, and above all, the capability to ultimately get it done.
• “Cold Shutdown” is used as a magic word. Public expects this would drastically improve the radiological environment so that refugees may be able to go home soon after this event. But this will never become true because contamination already there will continue to stay there regardless of the state of reactors.
33
Issue – Technical Perspectives• Applying BWROG’s EPG after having significantly deviated fro
m it currently creates a major “side effect”, that is, generating large volume of highly contaminated water. The longer they operate “Cold Shutdown”, the more contaminated water they generate and partly leak to the external environment (groundwater and seawater).
• Cannot apply it for Unit 2 because of major defect caused on the pressure boundary of its Suppression Chamber.
• Cannot cool the vicinity of “gas pocket” at the top of Skirt. The NRC report (NUREG/CR-6402 Rev.2) had pointed out a possibility of delayed creep rupture.
34
Delayed wall creep rupture would eventually occur in the vicinity of gas pocket.
35
Questions
• Is “Cold Shutdown” a mandatory milestone even 2 months after cease of chain reaction?
• Is there any better approach to manage the residual heat which is now only ~0.1% of the rated thermal power?
• How worse could it be at all if “Cold Shutdown” is abandoned? It has never been achieved to date for a long time anyway. What is the reason to have to stick to it?
• Achieving “Cold Shutdown” is OK. But, then what? – How long to keep it?
– Dismantling reactors, next?
– Back to “Green Field” eventually?
36
Difficulties to Dismantle BWR Reactors with Damaged Core
Reasons why so difficult• Requires full restoration of Refuel Floor, a part of Secondary C
ontainment and Reactor Building itself along with its ventilation system, as well as Overhead Crane, Fuel Handling Machine and all other special service tools prior to disassembly of RPV Head.
• Flooding RPV would result in leakage from the Bottom Head.• Removal of RPV Head is a challenging task.
– Studs/Nuts potentially severely galled.
– Radiation level too high to install/operate Stud Tensioner.
• Removal of Steam Dryer is more challenging due to high radiation level.
37
Reasons why so difficult (cont’d)• Removal of Moisture Separator is even more challenging.
– Shroud Head Bolt possibly galled due to exposure to elevated temperature and cannot be unlatched by following conventional procedure. Possibly requires to sever by remote EDM.
• Removal of Fuel Assembly is yet more challenging if not impossible. Most, if not all, FAs could have been fully destroyed, deformed and fused each other. Most part of core now possibly locates below Core Plate as “core debris”.
• Complete retrieval of once-molten core debris below Core Plate requires an exhaustive effort.
• Once core debris exits RPV and flows into the pedestal region, further efforts are not even theoretically possible unless Primary Containment Vessel is totally flooded.
38
Steam Dryer
39
Moisture Separator
40
Fuel Assembly
41
Lower Plenum (Region below Core Plate)
42
Justification to pursue• Efforts to place nuclear material under better inventory control
are in line with IAEA requirements for security reason.• Efforts to contain nuclear material within certified containers as
much as possible are considered more ethical practice.• Possible contribution to reduce long term risks associated with
decay heat and radio-toxicity. • Opportunity to gain detail technical data to be shared with
international community to improve management of severe accident. (e.g. Improving accuracy of analytical codes.)
Justification not to pursue• Too much technical and financial uncertainties to pursue.• Possibility to reduce short term risks associated with safety
and security. (e.g. Exposing damaged reactors to unsecured condition for extended period not preferable.)
43
Conclusions• Dismantling BWR Reactors with damaged core is technically
extremely challenging due to harsh radiological environment and not fully achievable any way.
• “Liquidation” strategy with damaged reactor left as is should be considered as one of the practical options.
• However, such a shortcut option requires public support domestically (including local communities) and internationally. Concurrence from IAEA may be necessary to comply with requirements for an NPT member country like Japan.
44
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45
Difficulties to retrieve spent fuel from degraded Reactor Bldg.
Why difficult?• Currently no place to go and no certified container to load for t
he damaged fuel assemblies. Therefore, fuel inspection (sipping) or other technically acceptable method to separate damaged fuel from intact fuel is necessary step to proceed.
• Fuel Handling Machine (FHM) is needed for fuel inspection.• Overhead Crane (OHC) with a loading capacity greater than 100
MT is needed for lifting spent fuel casks.• Both FHM and OHC are not currently functionable. Restoration
of OHC requires a major work to repair degraded Reactor Bldg.
46
Why difficult? (cont’d)• Using other type of crane or lifting machine is an optional
choice. However, it needs to be certified for safety application when handling spent fuel casks.
• New analyses and/or experiments may be required for the existing cask designs to be used.– All existing cask designs have been certified to meet requirements
to survive a set of design basis accidents including 9-meter free drop. However, using other crane may allow exceeding this condition.
– “Fuel-Handling Accident” (accidentally dropping one fuel assembly over the core) has been assumed as one of the design basis accidents for safety analysis and licensing both in Japan and the US. Standby Gas Treatment System has been assumed available under this postulated accident.
– “Spent Fuel Cask Drop Accident” is another design basis accident considered in the US for licensing, but not in Japan.
47
48
Justification to pursue• Efforts to place nuclear material under better inventory control
are in line with IAEA requirements for security reason.• Efforts to contain nuclear material within certified casks as
much as possible are considered more ethical practice and to be exercised wherever reasonably possible.
• Possible contribution to reduce long term risks associated with decay heat and radio-toxicity.
Justification not to pursue• Too much financial hardship to pursue.• Possibility to reduce short term risks associated with safety
and security. (e.g. Exposing physically unprotected SFP to unsecured condition for extended period not preferable.)
49
Conclusions• Retrieving SNF from degraded Reactor Bldg. is technically
challenging under existing licensing scheme especially when FHM and OHC are not available, and without intact Secondary Containment as well as emergency ventilation system (SGTS).
• “Liquidation” strategy with all SNF left in the existing SFP should be considered as one of the practical options for the units where FHM, OHC, Secondary Containment, and SGTS are lost.
• However, such a shortcut option requires public support domestically (including local communities) and internationally. Concurrence from IAEA may be necessary to comply with requirements for an NPT member country like Japan.
50
Intentionally left blank
51
Conventional Decommissioning Processes
Typical Decommissioning Processes in the US• Experiences:
– PWR: Yankee Rowe, Haddam Neck, Maine Yankee, Trojan
– BWR: Big Rock Point
• Goal: back to “Green Field”• Licensing Process:
– PSDAR (Post Shutdown Decommissioning Activity Report)
– LTP (License Termination Plan)
• Spent Fuel: Loaded in Dry-Cask and stored at site (ISFSI) until federal government determines what to do.
• Major Impacts Previously Experienced:– Inflation of Disposal Cost
– Significant Soil Contamination
52
Maine Yankee Experience• Reactor Type/Size: Three-Loop PWR, 2,700MWt/860MWe• Operational History: 12/28/1972 – 12/06/1996 (24 years)• Decommissioning Activities:
– Facility Demolition: 1997 – 2005
• Original Cost Estimate (as of 1997): $380M
• Overrun Cost: $26.8M
– Removal of Spent Fuel: Aug. 2002 – May 2004
• Original Cost Estimate (as of 1997): $128M
• Overrun Cost: $6.8M
– Personnel Exposure:
• Original NRC estimate: 11,150men-mSv
• Actual Result: Approx. half of estimate
– Techniques Applied:
• Chemical Decontamination
• Implosion
• Underwater High Pressure Abrasive Water Jet Cutting
53
Maine Yankee Experience (cont’d)• Decommissioning Activities (cont’d):
– Radioactive Waste:• Total Amount: ~140,000ton
• Amount transported to storage site: 88,450ton (63% of total) mostly by train
– Radioactive Waste (Reactor Internal Components)
Weight Activity
Ton % Bq %
Loaded and shipped in RPV 116 70 0.15E16 2
Loaded and shipped in casks 33 20 1.09E16 15
Loaded in GTCC and stored at site (4 GTCC casks)
16 10 6.03E16 83
Total 165 100 7.26E16 100
54
86% completion as of April 14, 2004
55
Implosion on Containment Bldg., September 17, 2004
97% completion as of January 19, 2005
56
Essentially 100% completed, as of May 5, 2005
“Green Field” achieved on July 25, 2005
57
Implosion Technique
Applied for Turbine Bldg.
58
ISFSI Pad and Spent Fuel Storage Casks
Vertical
Horizontal
59
Yankee Rowe Experience
ISFSI for Storage of 16 dry casks containing 533 spent fuel assemblies
Prior to Decommissioning Activities (1993)
Most Decommissioning Activities done (12/12/2006)
Decommissioning cost : $608M
600MWt PWR (1963 – 1991)
60
Back to “Green Field” as of 9/5/2007
Actual and Future Yankee Rowe Decommissioning Schedule
61
Cased in container on 11/20/1996
Departed from site on
4/27/1997
Loaded on to railcar for 1800km
transportation
Arrival at Barnwell Site
for subsurface
repository on 5/7/1997
Reactor Vessel Disposal
3.6m-dia. x 8.1m-tall, weighing 165tons
80 tons of concrete poured
inside and outside vessel
62
Large volume of subsurface soil found contaminated with tritium (H-3).
Numbers indicate H-3 concentration in groundwater in pCi/L.
EPA drinkable level is 20,000pCi/L.
63
Zion Project
3250 MWt PWR
Operational History
Unit 1 thru 1996 Unit 2 thru 1997
64
Basically just cooling-down
Dismantling Activities
65
30-year long project!
Finally Back to Green Field in 2028
SNF Disposition Campaign
66Source: NUREG-1350 Vol.21
Ultimate Solution for SNF if not Recycled
Yucca Mountain Project (abandoned) 500 to 600m deep geological repository
6767
Swedish Plan (active)
68
Intentionally left blank
69
Definition of “Plan-A”
• It has never been formally defined yet.
• So let’s assume that it is a recovery process, or decommissioning process to achieve “Green Field” as traditionally attempted in Japan (e.g. JPDR) and for several NPPs in the US.
• It has never been formally defined yet.
• So let’s assume that it is a recovery process, or decommissioning process to achieve “Green Field” as traditionally attempted in Japan (e.g. JPDR) and for several NPPs in the US.
70
Feasibility to apply US Decommissioning Experiences and Lessons-Learned for Fukushima NPP
• Demolition Techniques:– Some applications possible but scope limited.
– Hindered mostly due to high level radiation/contamination.
• Site Restoration:
“Back to Green Field” is practically an impossible goal.– Wide spread contamination: Soil, Groundwater, Seawater
– Highly radiotoxic actinide species (Pu) involved.
– No candidate repository locations available for large volume of heavily contaminated equipment and concrete rubble.
Conclusion: “Plan-A” is not a workable option for Fukushima NPP Units.
71
Unit 1 2 3
mSv/h 46,500 18,200 8,400
Technique Application for Fukushima NPP Units
Implosion
Possible for all buildings other than Rx. Bldg. of Unit 1 to 3 after some decontamination efforts.
Not practical for Rx. Bldg. of Unit 1 to 3 due to high contamination level.
Chemical DecontaminationAlready done for Unit 4 Rx.
Not practical for other units due to too much activity load.
Remove Rx. Internals by High Pressure Abrasive Water Jet
Already done for Unit 4.
Not practical for other units due to high contamination level.
Separate RPV from All Other Connecting Systems
Possible for Unit 4.
Not practical for other units due to harsh radiological environment for workers.
(Drywell Dose Rate as of May 20, 2011.)
Note that the dose limit for Emergency Workers is
250mSv.
72
Intentionally left blank
73
Developing “Plan-B”
• Worst Case Scenario– Historical worst case: Chernobyl Accident– How better is Fukushima NPP’s case relative to Chernobyl?– How worse could Fukushima NPP units have been if cooling
capability was lost immediately upon SBO?– How bad could Fukushima NPP units be if “Cold Shutdown”
is terminated now?• Unit 1/2/3 Reactors• Unit 4 SFP
• Basic Technical Requirements for “Plan-B” • Transition from “Plan-A” to “Plan-B”
74Source: NUREG/CR-6042 Rev.2
Worst Case Scenario• Historical Worst Case: Chernobyl Accident
Fukushima NPP
NISA: 1,700,000Ci
NSC: 1,000,000Ci
75
Chernobyl Accident Facts:
Definitely worst from many aspects!
• Release: 1,760PBq of I-131• Contamination:
– 3 major “Hot Spots”, including one as far as 500km from the site.
– Large restricted areas
confiscated zone, closed zone, permanent control zone, periodic control zone
• Personnel Exposure (out of 400 workers at the site on the day of accident):– 140 persons 1 – 2 Gy
– 55 persons 2 – 4 Gy
– 21 persons 4 – 6 Gy
– 21 persons 6 – 16 Gy
76
Radionuclide Releases During Chernobyl Accident
Source: Chernobyl – Ten Years On (OECD/NEA)
77Source: NUREG/CR-6042 Rev.2
1MC
i = 3
7,00
0TB
q
Daily Release During Chernobyl Accident
78
Source: OECD/NEA “Chernobyl Ten Years on Radiological and Health Impact – An Assessment by the NEA Committee on Radiation Protection and Public Health” November 1995
Cs-137 Contamination 10 years later
Vicinity of Fukushima NPP
80km
Equivalent dose rate of 555kBq/m2 contamination is 1.8μSv/h or 15.8mSv/y. Blue colored r
egion on land represents dose rate greater than 0.3μSv/h as of 3/19/2011.
500km
79
Access Restriction due to High Level Contamination
80
Worst Case for Fukushima NPP
• In spite of large amount of release, resulting overall impact was much smaller than that of Chernobyl accident. This is believed to be mostly because of wind blowing west to east. The worst case was avoided by a favorable wind direction.
Prediction by WeatherOnline (UK)
81
Radiological impact estimated by various organizations
Japan - estimated cumulative dose in mSv through 3/11/2012
82
83
Worst Case for Fukushima NPP (cont’d)
What if cooling capability was lost immediately upon SBO?
• Much more heat load, resulting in more aggressive propagation of failures/degradations of FP barriers. (Reactor Pressure Vessel, Primary Containment, and even Man-Made Rock. Note that Man-Made Rock is not credited as an FP barrier against atmospheric release, but it does play a role as an FP barrier against groundwater/soil contamination.)
• Earlier and much more release of radionuclides, with greater contribution from short-lived species.
• Much more CCI, resulting in generation and accumulation of more non-condensable and combustible gases forming radionuclide aerosols containing more radiotoxic particles.
84
Unit 1 2 3 4
# of Fuel Assembly in Rx. 400 548 548 0
Electrical Output (MWe) 460 784 784 0
Thermal Output (MWt) 1,380 2,381 2,381 0
Estimated Residual Heat (MWt)1% of rated Thermal Output
14 24 24 0
Residual Heat Generation
2 months after
shutdown1 hour after
shutdown
85
Major Degradation of Reactor Pressure Vessel Bottom Head and Core-Concrete Interaction (CCI), Resulting in Significant Amount of Release of Radioactive Aerosol
Pedestal Doorway
Pedestal
H2O,
CO2
H2O,
CO2
H2, CO
Aerosol
AerosolAerosol
Aerosol
86
Pedestal DoorwayPedestal
Aerosol
Aerosol
Aerosol
Aerosol
H2, CO
H2O,
CO2
H2O,
CO2
Beginning of Primary Containment Melt-Through
87
Gross Failure of Primary Containment due to Melt-Down Progression
Aeros
ol
Aeros
ol
Aeros
ol
Aeros
ol
88
Gross Man-Made Rock (Basemat) Melt-Through
Man-made Rock
89Source: NUREG/CR-6042 Rev.2
Various Gases and Debris Generated during CCI
90Source: NUREG/CR-6042 Rev.2
Breakdown of FP Species
91
Worst Case for Fukushima NPP (cont’d)
How bad could Fukushima NPP units be if “Cold Shutdown” is terminated now (70 days after shutdown)?
• Much less heat load, resulting in less aggressive propagation of failures/degradations of FP barriers (Reactor Pressure Vessel and Primary Containment).
• Much less release of radionuclides, with negligible contribution from short-lived species.
• Much less CCI, resulting in generation and accumulation of less non-condensable and combustible gases forming radionuclide aerosols containing less radiotoxic particles.
• Water left on the Drywell floor suppresses CCI. • Dilution of molten core with various metal and non-metal materi
als would lower temperature and reduces CCI.
92
Radioactive Decay after 70 days
Half-Life Remaining
1 day 8.5 x 10-22
2 days 2.9 x 10-11
3 days 9.5 x 10-8
I-131,
Cs-134, Cs-136, Cs-137, Rb-86,
Te-127m,
Ba-140, Sr-89, Sr-90,
Co-58, Co-60, Ru-103, Ru-196,
Am-241, Cm-242, Cm-244, Nb-95, Nd-147, Pr-143, Y-91, Zr-95,
Ce-141, Ce-144, Pu-238, Pu-239, Pu-240, Pu-241
32 species gone, 27 species left.
93
Gross Failure of Primary Containment Melt-Through
This would be very slow even if it does take place at all.
94
Complete Melt-Down through Man-made Rock (Basemat)
This is even more unlikely. 12~
13m
95
Intentionally left blank
96
Basic Technical Requirements for “Plan-B”
• Achievable and Predictable– No optimistic assumptions allowed. (e.g. Equipment inside Drywel
l no longer functionable.)
• Designed for Short-term and Long-term Solution– Simplicity and Passiveness
– Good for 1,000 years with minimum maintenance.
– Maintain reversibility in case application for long-term solution is abandoned in the future.
• Short-term and Long-term Safety/Security– No Recriticality, H2/Steam Explosions
– Minimum release of radioactivity to external environment
• Lowest Cost and Shortest Schedule– Least labor intensive, minimum personnel exposure.
97
Transition from “Plan-A” to “Plan-B”
• Water Cooling to Gas/Air Cooling (Forced Circulation to Natural Circulation)
• Proposed Remedy for Leaky System– Introduce fine glass fiber mixed with SiC/B4C powder to clog leak p
aths, then use some chemical reaction to create precipitants (e.g. Ca3(PO4)2 ) to further reinforce leaking boundary.
– Apply knowledge gained from GSI-191 study.
• Proposed improvement to minimize I-131 airborne inside and outside Rx. Bldg if such an effort is still considered necessary.– Spray TSP (Trisodium Phosphate) solution.
98
Intentionally left blank
99
“Plan-B”
• Liquidation Strategies for Fukushima NPP Reactors and SFPs– Strategy-I and II for Reactors– Strategy-A and B for SFPs
• Water Treatment and Entombment• On-Site Above-Ground Repository• Design Beyond Millennium• Beyond “Liquidation”
100
Liquidation Strategies for Fukushima NPP Reactors and SFPs
General Approach• New concept replacing traditional and costly practice, that is, full
demolition of the facility and eventually returning the entire site to “Green-Field”. With this new approach, most radioactive material is left as is in the original locations so that significant cost/schedule reduction is expected.
• Thoroughly engineered design, not like the ad hoc technique applied for Chernobyl Unit 4 under emergency situation.
• Applicable not only for those plants affected by major reactor accidents, but also for those plants orderly shutdown permanently upon expiration of license as an alternative choice.
• Phased approach to shift cooling strategies as decay heat load decreases as a function of time . Forced circulation (Water to Helium, Helium to Air) initially, eventually followed by natural convection with no active component to drive the system.
101
General Approach (cont’d)
• This new approach is named as IE2-D (Innovative EngineeredIn-Situ Entombment Decommissioning) and comprised of the following general decommissioning processes and specific processes unique to each Strategy described in later sections separately:
– Remove all new fuel assemblies currently stored in the New Fuel Storage Vault.
– Remove any equipment reasonably recyclable.
– All process systems containing water inside are to be drained, filled with N2 gas and isolated from external environment.
– All rooms and compartments are either solidly filled with concrete, or vented to the general area so as not to allow accumulation of combustible gas.
Unit 1 2 3 4
# of New Fuel Assy 100 28 52 204
102
Specific Strategies
Unit Reactor Spent Fuel Pool
1 Strategy-II Strategy-B
2 Strategy-IIStrategy-A
SFP Not Affected
3 Strategy-II Strategy-B
4Strategy-I
Reactor Systems Not Affected
Strategy-B
103
Strategy-I
Status
• Implementable only for Unit 4 because this is the only unit where most reactor systems are apparently left unaffected.
• However, the unit was structurally significantly damaged due to the hydrogen explosion, resulting in losses of OHC and possibly FHM as well.
• As another impact due to the hydrogen explosion, the integrity of secondary containment is currently lost.
Unit 1 Unit 2 Unit 3 Unit 4
104
Status (cont’d)
• The unit was in the middle of outage where a major modification project, namely “Shroud Replacement”, was taking place at the time of accident. The reactor configuration during this particular outage was very different from that during normal outages, specifically;
– Steam Dryer and Moisture Separator were removed from the vessel and stacked together vertically in the Dryer Separator Pit.
– All Control Blades, Fuel Support Castings, Control Rod Guide Tubes, and Incore Monitors were removed from the vessel and temporarily stored in the SFP.
– Many reactor internal components including, Feedwater Spargers, Core Spray Piping and Spargers, Top Guide, Core Shroud, and Core Plate were removed from the vessel and transferred to the Dryer Separator Pit where some of them were sliced into small pieces for disposal.
105
Key Steps
• Inspect the Reactor Bldg. and determine degree of impact.
• Clean up Refueling Floor.
• Restore and re-establish capability of Secondary Containment.
• Move all removed reactor internals currently stored in the Dryer Separator Pit and SFP in an orderly manner back to the vessel.
• Fill the Reactor Pressure Vessel with concrete.
• Drain Reactor Cavity and Dryer Separator Pit. (These emptied pools will be used for the storage of various contaminated equipment and debris for the future.)
• Re-assembly RPV Head, Mirror Insulation, PCV Head.
• Proceed to the general decommissioning processes.
106
Intentionally left blank
107
Strategy-II
• Proposed Unit-by-Unit Application– Mode 1, 2 (Options A, B1, B2, C)
– Mode 3
• Tentative Mode Change Schedule• Proposed System Lineup
Detail plant unique assessment is necessary.
Unit 1 Unit 2 Unit 3 Unit 4
108
• Proposed Unit-by-Unit Application
UnitHeat Load*
(kWt)Operation Mode**
Min. Flow Rate*** (Nm3/h)
1 1,400
1 He, Forced 55,000
2 Air, Forced 15,000
3 Air, Natural 5,000
2/3 2,400
1 He, Forced65,000
(∆T = 150 deg-C)
2 Air, Forced 15,000
3 Air, Natural 5,000
*: Residual heat generation as of 5/11/2011.
**: See conceptual illustrations and proposed system lineup for each operation mode.
***: Required flow rate is calculated to limit the outlet temperature within 100 degrees above the inlet temperature unless otherwise noted.
109
Operation Mode 1 2 3
Heat Generation Range (kW) > 700 200 - 700 < 200
Cooling Strategy He/Forced Air/Forced Air/Natural
Estimated Heat Generation as of 5/11/2011
Unit 1 1,400kW
Unit 2 2,400kW
Unit 3 2,400kW
Unit 4 0 N/A
Mode 1
Mode 2 Mode 3
Mode 2 Mode 3
1Y 2Y 10Y
Mode 1 Mode 2
Mode 1
3Y
Mode 3
• Tentative Mode Change Schedule
110
Medium
Thermal Conductivity
Thermal Conductivity
Heat Capacity
W/m ・ K Air = 1 J/kg ・ degC
He 0.1663 5.53 5192
H2O (Steam) 0.0241 0.77 2098
H2O (Liquid) 0.582 - 4217
Air 0.0316 1 1012
Helium:
• 200Yen/Nm3
• 140Yen/liter (liquid)
• 0.1248kg/liter
Helium:
• 200Yen/Nm3
• 140Yen/liter (liquid)
• 0.1248kg/liter
Favorable Thermal Characteristic of Helium
111
Helium is a standard cooling medium for high temperature gas reactors.
GT-MHR (Gas Turbine – Module Helium Reactor)
112
A*
B*
To be added
Scrubber/Gas Cooler
Ventilation System
Mode-1/2
Heat Sink Gravel
Flow from Suppression Chamber
to Drywell
*: See “proposed line-up” for system interfaces for A and B for each unit.
Option A
113
Heat Sink Gravel
Factors to be considered for selection:
• High thermal conductivity
• Radiation shielding
• High performance to absorb radioactive gas/particle.
114
Copper Sphere ShellZeolite
Mixing several different constituents may be considered
115
Flow from Suppression Chamber to Drywell
116
Field Assembly of Primary Containment at Browns Ferry Site During Construction Time
117
Unit 1 Core Spray SystemHelium/Air Injection Point
Proposed System Lineup
A
118
Unit 1 Shutdown Cooling System
B
This valve may not be opened.
119
Unit 1 Isolation Condenser (Alternative Option)
B
X
120
Unit 2/3 Core Spray SystemHelium Injection Point
A
121
Unit 2/3 High Pressure Injection System
B
122
Unit 1 Atmospheric Control System (Alternative Option)
B
X
123
To be added
To be added
Scrubber/Gas Cooler
Ventilation System
Rx. Bldg. Truck Bay
Mode-1/2
Option B1
124
To be added
To be added
Scrubber/Gas Cooler
Ventilation System
Rx. Bldg. Truck Bay
Blower
Mode-1/2
Option B2
125
To be added
To be added
Scrubber/Gas Cooler
Ventilation System
Rx. Bldg. Truck Bay
Mode-2
Option C
126
Rx. Bldg. Truck Bay
StackAir Gap for Flow Path
Air Flow only by Natural Convection
Mode-3
See detail “D”
See detail “E”
127
Detail “D”
128
Detail “D”
129
Construction Details of Bottom Portion of Primary Containment Vessel (Oyster Creek) (2)
Detail “E”
130
Construction Details of Bottom Portion of Primary Containment Vessel (Oyster Creek) (1)
Detail “E”
131
Intentionally left blank
132
Strategy-A
• Implementable only for Unit 2 because both OHC and FHM are seemingly still functionable and available.
Key Steps:• Inspect (sipping and visual examination) on all fuel assemblies
and identify any damage fuel.• Load only undamaged/non-degraded fuel assemblies into cask
s for:– Wet Storage at Common Storage Pool or other designated site(s).
– Reprocessing for MOX at Rokkasho facility.
– Dry Storage at site or other designated site(s).
Unit 1 Unit 2 Unit 3 Unit 4
133
• Damaged/degraded fuel assemblies are treated differently.– No certified cask design currently available.
– Design and certify special cask only for this group of fuel and transfer to other unit (1, 3, or 4) for Strategy-B.
– Leave only this group of fuel at Unit 2 and apply Strategy-B.
134
Intentionally left blank
135
Strategy-B
• Proposed Unit-by-Unit Application– Mode 1, 2, 3
• Comparison with “Plan-A”
Advantages vs. Disadvantages– General Comparison– Cost– Schedule– Security during implementation
Unit 1 Unit 2 Unit 3 Unit 4
136
• Proposed Unit-by-Unit Application
UnitHeat Generation
(kWt)Recommended
Operation Mode*Min. Flow Rate**
(Nm3/h)
1 70 3 Air Natural 1,930
2 4601 He***, Forced 18,000
2 Air, Forced 13,000
3 2302 Air, Forced 6,500
3 Air Natural 6,500
4 1,800 1 He***, Forced 70,000
*: See later section for the definition of each operation mode.
**: Required flow rate is calculated to limit the outlet temperature within 100 degrees above the inlet temperature.
***: He is more recommendable because of its higher heat conductivity and lower viscosity (flow friction).
137
Operation Mode 1 2 3
Heat Generation Range (kW) > 350 100 - 350 < 100
Cooling Strategy He/Forced Air/Forced Air/Natural
Estimated Heat Generation as of 5/11/2011
Unit 1 70kW
Unit 2 460kW
Unit 3 230kW
Unit 4 1,800kW Mode 3Mode 2Mode 1
Mode 1
Mode 2 Mode 3
Mode 3
Mode 2 Mode 3
0.5Y 2Y 5Y 5.5Y 10Y
• Tentative Mode Change Schedule
138
N
Spent Fuel Racks Spent Fuel Racks
Gate
Cask Pit
A A
Spent Fuel Pool (top view)
139
Water Level
Spent Fuel Racks Spent Fuel Racks
Spent Fuel Pool (side view)
A-A
140
Finned Heat Sink Chambers (Copper)
Step-1 Install Finned Heat Sink Chambers on Spent Fuel Racks.
141
Cross-Tie Pipes
142
Step-2 Install Pre-fabricated Pipe Modules.
143
35cm
Cold (Inlet) 2-inch Sch#40 Stainless Steel
Hot (Outlet) 2-inch Sch#40 Stainless Steel
Convection Cooling 2-inch Copper
Pipe Modules
144
eachcm,xcmAvailableAreaFlowTotal
each.approxInstalledTubesofNumber
m)SurfaceTopEntireof%(~AvailableAreaTotalAssumed
cmCellUnitofArea
22
2
22
9001890021
900
5050
530354
3
Φ10mm (typ. 4)
50m
mA
pp
rox.
200
0mm
Ap
pro
x. 8
000m
m
145
A A
View A-A
Main Header
Main Header
Main Header
Distribution Header
Top View
146
147
Step-3 Load Heat Sink Gravel
Water
Gravel
148
Heat Sink Gravel
Factors to be considered:
• High thermal conductivity
• Radiation shielding
• High performance to absorb radioactive gas/particle.
149
Copper Sphere ShellZeolite
Mixing several different constituents may be considered
150
Water level gradually decreases
Step-4 Start Ventilation System
151
To Ventilation Fan and Gas Treatment System
Wet Scrubber
Water level
Operation Mode 1, and Mode 2
152
Mode Heat Load Cooling MediumCleanup System
Power
1High
(>350kW)
Contaminated Helium
Required Forced Cooling
2Medium
(100-350kW)
Contaminated Air
Required Forced Cooling
3Low
(<100kW)
Non-contaminated Air
Not RequiredNatural
Convection
Operation Mode
Medium
Thermal Conductivity
Relative Thermal Conductivity
Heat Capacity
W/m ・ K Air = 1 J/kg ・ K
He 0.1663 5.53 5192
Air 0.0316 1 1012
Favorable Thermal Characteristic of Helium
153
Operation Mode 3 “Natural Convection”
Inlet Sleeve
Shielded Air Intake Block
154
155
Comparison with “Plan-A”
Advantages vs. Disadvantages• General Comparison
– Favorable for “Plan-B”
“Plan-A” “Plan-B” Strategy-B
FHMRequired but currently not available due to damage caused by H2 explosion.
Not required.
OHC Ditto Not required
Fuel Inspection (Sipping)
Required Not required.
Spent Fuel StorageNo certified Transportation/Storage Cask for damaged fuel assemblies currently available.
Not required.
ISFSI ?? Not required
SecurityCurrently exposed to very poor conditions.
Duration of poor security conditions can be minimized.
156
Advantages vs. Disadvantages (cont’d)• General Comparison
– Potential issues associated with “Plan-B” Strategy-B.
“Plan-A” “Plan-B” Strategy-B
Geological DisposalCan be eventually transferred to this option.
• Practice not pursued previously.• Buried under man-made structure significantly above ground elevation.
Licensing ProcessRelatively more predictable.
• Unknown. No previous experience. • No siting criteria established.
Safety AnalysisRelatively more predictable.
Various supporting analysis necessary.• Design Basis Accident (DBA)
Security Issue Currently very poor.Permanent measures including Aircraft Impact Assessment (AIA) necessary.
Public Acceptance Unknown. Unknown.
157
• Cost/Schedule– Cost/schedule potentially eliminated by applying “Plan-B”
Strategy-B.
– Cost for “Plan-B” Strategy-B: much less than that for restoring OHC alone.
– Schedule for “Plan-B” Strategy-B: much shorter than that for unloading spent fuel from SFP alone.
Activity Cost (JPY) Schedule (Year)
Restore OHC X billion ~2
Restore FHM
Fuel Inspection (Sipping)
Procure Spent Fuel Casks X billion
Unload Spent Fuel from SFP ~5
Construct ISFSI
158
Conclusions:• Practical approach for Units 1, 3 and 4. (Strategy-A is considered
implementable only for Unit 2.)
• Advantage of “Plan-B” Strategy-B over “Plan-A” is obvious.
• All associated technical issues are manageable.
• Two potentially challenging non-technical issues:
– Licensing
– Public Acceptance
159
Ultimate Configuration with Operation Mode 3
All contaminated equipment and materials are permanently buried in-situ.
Paradigm Shift !!
This concept, in spite of huge cost benefit expected, significantly deviates from the conventional approach.
Paradigm Shift !!
This concept, in spite of huge cost benefit expected, significantly deviates from the conventional approach.
160
Intentionally left blank
161
Water Treatment and Entombment
• Water Treatment System is a part of “Plan-B” and integrated into IE2-D strategies.– Low level contaminated water is used as a water source to
produce ready-mixed concrete for general purpose.– Highly concentrated radioactive water is vitrified (because
of relatively high heat generation) and stored at On-Site Repository.
• Water to be processed:– Highly contaminated water currently stored in various pools
at site.– Contaminated sea water within Intake Area.
162
Water Treatment (1)
Highly Contaminated Water Currently
Stored in Various Pools at Site
Vitrification
Canisters
On-site Repository
Concentrated Radioactive Liquid
Treatment System
Cement
Aggregate
Contaminated Concrete Rubble (Optional)
Processed Water (slight contamination
allowed)
Ready-Mixed Concrete
< 5,000Bq/cm3
< 0.065mSv/h
20 v/v% 80 v/v%
For Entombment Work
For Entombment Work
163
Dose Rate Calculation of Homogenously Contaminated Concrete
h/mSv.or,h/SvR
hBq/Svx.
e.
R
m/BqQ
eQ
dreQR
m.r
er
QdrrdR
.x.
.r. r
r
065065
10928
12517
104
10
44
828
14
8
82825179
39
828
0
828
0
22
Assumption:
500TBq in 105 m3, or 5,000Bq/cm3 of processed water
Water Content in Ready Mixed Concrete = 20%
Calculation:
Low enough!
164
Heat Generation Calculation of Homogenously Contaminated Concrete
Assumption:
500TBq in 105 m3 of water, or 5,000Bq/cm3
Water Content in Ready Mixed Concrete = 20%
Energy Release per Disintegration = 1MeV
Calculation:
Total energy release rate = (1.6 x 10-13J) x (5 x 1014/sec) = 80W
Temperature increase based on black body radiation
q” = σT4 σ = 5.67 x 10-8
q” = 80/(4πr2) r = 28.8m
q” = 7.7 x 10-3 W/m2
T = 19-deg C
Low enough!
165
Water Treatment (2)
Desalination System
Cement
Aggregate
Contaminated Concrete Rubble (Optional)
Processed Water (still slightly
contaminated)
Ready-Mixed Concrete
< 5,000Bq/cm3
< 0.065mSv/h
20 v/v% 80 v/v%
Contaminated Water within Intake Area
For Encasing Concrete Rubble
For Encasing Concrete Rubble
166
Intentionally left blank
167
On-Site Above-Ground Repository
• New site arrangement consists of three major islands, each enclosed by an individual protected area:– ISFSI– On-Site Repository for vitrified canisters– IE2-Ded Reactors
168
Intake Area
Intake Facility (typ.)
Backwash Valve Pit (typ.)
Control Bldg. (typ.) Turbine
Bldg. (typ.)
Rx. Bldg. (typ.)
RW Bldg. (typ.)
BeforeUnit 2
Unit 1Unit 3Unit 4
169
Entombed ReactorsAfter
Protection Fence against Aircraft Impact
Stack
Concrete rubble generated from demolition of all other structures is encased in the large concrete block(s).
Tsunami Barrier
Tsunami Barrier
Wave Breakers for Tsunami Protection
Original Shoreline
170
ISFSI for SNF and any potential GTCC Waste
On-Site Repository for Vitrified Canisters
Legend:
Monitoring Post
Ground Water Sampling Point
Protected Area
Main Gate
New Site Boundary
Monitoring Facility
Conceptual New Site Arrangement
171
Intentionally left blank
172
Design Beyond Millennium
• IE2-Ded Reactors, ISFSI, and On-Site Repository must be qualified for long-term endurance.
• Traditionally, man-made structures were not credited for this purpose.
173
But, man-made structures may not be too bad…
Possibly good for centuries or even millennia!
174
Source: “The Future of Nuclear Power” (MIT)
Residual Heat
1/20
175
Source: “The Future of Nuclear Power” (MIT)
Radioactivity
1/100
176
Intentionally left blank
177
District for New Industry/Community Development
Entombed Reactors (Units 1 to 4)
Survived Reactors (Units 5 and 6)
Solar Thermal PowerBeyond “Liquidation”
Previous Site Boundary
178
Target Overall Schedule
Activities 2y 4y 6y 8y 10y
Public Acceptance (Workshop)
Licensing Review on EI2-D (Safety Analysis)
Build Liquidators’ Villages
Recruit Liquidators
Expand On-Site Liquidation Infrastructures
EI2-D Projects
Unit 1 to 3, Reactor
Unit 4, Reactor
Unit 1, 3, and 4, SFP
Unit 2, SFP
Demolition of other structures
Construct On-Site Repository Facility
Water Treatment, Vitrification
Construct Intake Area Tsunami Barriers
Construct ISFSI (for Unit 2 SNF)
New Industry/Community Development
FS, Bidding, Design/EngineeringMode 1 Mode 2 Mode 3
Mode 1 Mode 2 Mode 3
Sipping Transportation Campaign
179
Intentionally left blank
180
Next Step
• Feasibility study by independent organization(s):– Technical aspect
– Financial aspect
– Political aspect
• Survey on public opinions.• Voices from international communities.
• Issues:– Interactions with decision-makers
– Financial support to proceed to design phase
– How to make a go-no-go decision
181
Analysis for Future Benefit
• Better predictability and versatility
182
Time Cooling Efforts Abandoned after Plant Shutdown
Release (NG, I, Cs)
CCI Penetration Depth
Required Evacuation Radius
183
184
Temperature Monitoring Probes
185
To be added
Rx. Bldg. Truck Bay
186
187
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