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Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-AC04-94AL85000. SAND2014-18957 PE
Sampling and Analysis of Dusts/Salts from In-service Storage Canisters at Calvert Cliffs, Hope Creek, and
Diablo Canyon ISFSIs Charles Bryan and David Enos, Sandia National Laboratories
SNL/BAM workshop, 8 October 2014, Albuquerque
10 µm
Overview • Background and ISFSI’s sampled
• Types of samples and sampling methods
• Sampling issues
• Analysis methods
• Hope Creek
• Diablo Canyon
2
Background • Stress corrosion cracking (SCC) of
stainless steel due to deliquescence of chloride-rich salts on the metal surface is well-known, especially in near-marine environments.
• Many Independent Spent Fuel Storage Installations are at coastal sites. Possible risk of SCC.
3
• EPRI sampling program: Assess the composition of dust on the surface of in-service stainless steel SNF storage canisters, with emphasis on the deliquescent salts.
• ISFSI locations sampled: • Calvert Cliffs: Transnuclear NUHOMS system,
horizontal storage canister (June, 2012) • Hope Creek: Holtec HI-STORM system, vertical
canister (Dec, 2013) • Diablo Canyon: Holtec HI-STORM system (Jan
2014) • Samples delivered to Sandia National
Laboratories for analysis
Calvert Cliffs NUHOMS system
Diablo Canyon HI-STORM
system
4
Calvert Cliffs Site
Eastern U.S.
ISFSI is ~0.5 miles from Chesapeake Bay • Sheltered bay • Brackish water
ISFSI
~0.5 miles
5
Hope Creek Site
Delaware Bay
Eastern U.S. ~15 miles
ISFSI ~0.25 miles
ISFSI is ~0.25 miles from the Delaware River, 15 miles upstream from Delaware Bay • Brackish water • Sheltered from
open ocean
6
Diablo Canyon Site
~0.35 miles
Western U.S.
ISFSI is ~1/3 mile from the shoreline, on a hill above the plant. • Elevated (~400 feet) above sea level • Rocky shore, breaking waves • Open ocean
Sampling Entry NUHOMS horizontal storage
systems—entered through door, annulus around shield plug (∼2.5 cm)
HI-STORM systems—entered through upper ventilation opening
Wet sampling Salt-Smart™ sensors Used to characterize soluble salts
(quantify amount per unit area) After use, sensor was split and
captured salts were rinsed out for analysis
Dry dust samples Scotch-Brite™ pads Used to characterize salt
components (chemistry, mineralogy, texture); cannot quantify amount per unit area
7
Area = 3 cm2
Salt-Smart™ sensor
Sampling a HI-STORM cask
Scotch-Brite™ pad
Analysis Methods SEM imaging and energy dispersive system (EDS) element maps
Dry samples Provide textural and mineralogical information Identification of floral/faunal fragments in dust
X-ray fluorescence Dry samples Micro-analytical technique—allows chemical mapping of the dry
pad surfaces with a resolution of ~50 µm Provides semi-quantitative chemical analyses—yields element ratios
that can be used in mass balance calculations Cannot detect elements lighter than sodium (e.g., oxygen, nitrogen)
X-ray diffraction Analysis of pads for mineralogical information
Chemical Analysis Dry pad and Salt-smart® samples leached with DI water, and the
leachate analyzed to determine soluble salts in the dust Insoluble fractions digested and analyzed to determine bulk chemistry
8
Results: Calvert Cliffs
9
100 µm
Dust particles adhering to a Scotch-Brite™ pad used to sample dust on the surface of an in-service storage canister, Calvert Cliffs ISFSI.
Dust on the storage canister surface, Calvert Cliffs ISFSI.
• The canister upper surface was more heavily coated with dust and salts due to gravitational settling. Samples from upper surface contained abundant pollen.
• The soluble salts are Ca- and SO4-rich. Gypsum is the dominant salt phase present.
• Chlorides comprise a small fraction of the total salt load, and are dominantly NaCl.
• Despite the proximity to the coast and prevailing winds from the east, the dusts sampled from in-service containers at Calvert Cliffs do not appear to have a large sea salt component. Chesapeake Bay is brackish, and may be sheltered sufficiently to limit wave-generated sea-salt aerosols.
100 µm
Pollen grains in dust on the upper surface of the canister.
Results: Hope Creek
10
Si O Mg
Fe Cr Ca
S Na Cl
N K Al
• Flat canister top much more heavily coated than vertical sides. • Dust dominated by insoluble minerals (quartz, clays, aluminosilicates). Soluble salts minor;
dominantly gypsum, carbonates. Sparse chlorides, mostly isolated grains of NaCl. • Despite the proximity to the coast, the dusts
sampled from in-service containers at Calvert Cliffs do not have a large sea salt component. Chesapeake Bay is brackish, and may be sheltered sufficiently to limit wave-generated sea-salt aerosols.
Chemistry: Hope Creek Salt-Smarts®
11
K Ca Mg Na NH4+ F– Cl– NO3
– PO43– SO4
2–
144-008 Side 13.0 93.2 0.8 3.4 0.6 0.1 2.7 0.9 2.7 4.1 15.4
144-009 Side 7.5 116.5 1.7 4.5 0.5 0.1 2.7 0.9 6.4 1.1 6.5 24.3
144-010 Side 1.0 133.9 1.4 4.2 0.4 0.4 2.4 1.2 5.0 4.4 19.4
144-013 Top 0.0 138 18 102 33 42 2.8 0.4 4.2 19 4.8 91 317
144-014 Top 0.0 141.2 6.4 29 8.0 13.4 2.7 0.4 18 7.3 1.3 55 142
144-003 0.6 2.2 0.4 1.4 0.5 3.3 1.2 2.1 11.6
144-004 0.3 3.2 0.6 2.9 0.8 1.8 0.5 1.7 11.8
145-006* Side 13.0 70.6 2.2 4.4 0.6 0.5 2.3 2.2 8.1 4.7 25.1
145-007 Side 7.5 100.8 1.0 2.4 0.6 0.7 2.9 2.1 2.2 0.7 5.3 17.9
145-014 Side 1.0 130.3 0.9 3.2 0.8 0.6 3.2 1.2 2.5 9.1 21.5
145-013** Top 0.0 174.1 15 91 30 32 2.8 2.2 15 3.5 82 273
145-011** 0.2 2.3 0.3 3.0 0.7 1.3 1.7 9.6
145-002 1.2 4.8 0.5 2.7 0.7 5.9 0.8 2.0 18.5
SS-Bl-8 min-1 1.3 0.2 1.1 0.4 1.6 0.6 5.1
SS-Bl-8 min-2 1.2 0.2 1.5 0.7 0.9 0.5 0.2 5.2
SS-Bl-15 min 1.5 0.5 5.7 0.2 0.7 1.1 1.6 1.7 12.9
* Pad only damp** Pad only partially saturated
SUMAmount present, µg/sample
Sample Loc. Depth, ft
Temp, ºF
Results: Diablo Canyon
12
• Canister sides lightly coated, tops heavily coated. • Dust dominated by insoluble minerals (quartz, clays, aluminosilicates), but chloride-rich soluble
salts are abundant, present as sea-salt aggregates. • Heavy wave action at the Diablo Canyon site
generates abundant sea-salt aerosols. Although 400 feet above sea level, Diablo Canyon canisters have a significant amount of sea-salts on the canister surfaces. Si O Mg
Al Ca S
Na Cl N
Fe K
Results: Diablo Canyon
13
• Canister sides lightly coated, tops heavily coated. • Dust dominated by insoluble minerals (quartz, clays, aluminosilicates), but chloride-rich soluble
salts are abundant, present as sea-salt aggregates.
Diablo Canyon Sea-salt Aerosols
14
Diablo Canyon Sea-salt Aerosols
15
Chemistry: Diablo Canyon Salt Smarts
16
Na+ K+ Ca2+ Mg2+ F– Cl– NO3– PO4
3– SO42–
123-003 Side 14.0 119.7 0.3 0.6 2.4 0.6 0.3 1.2 1.5 0.4 4.3 11.6
123-004 Side 11.5 173.4 0.2 1.2 2.6 0.4 0.1 0.9 3.7 0.1 2.1 11.4
123-005 Side 10.5 187.0 n.a. 0.3 3.6 0.2 0.3 0.5 0.6 0.5 1.4 7.2
123-002 — — — 14.4 0.9 6.0 0.9 0.9 14.1 11.3 n.a. 10.4 58.8
123-010 — — — 3.3 1.9 2.2 0.5 1.0 6.2 1.3 0.8 1.6 18.8
170-007 Side 10.5 177.5 1.0 0.3 2.0 0.3 0.3 1.0 1.9 n.a. 1.4 8.2
170-008 Side 9.5 182.8 0.2 0.5 2.4 0.2 0.3 0.7 2.3 0.6 0.6 7.9
170-009 Side 9.0 188.2 0.3 2.3 3.2 0.2 0.2 0.6 9.3 0.6 0.9 17.7
170-002 — — — 7.3 1.3 5.9 1.3 0.2 3.2 21.0 0.8 6.2 47.3
B1-6 — — — 0.7 0.9 1.8 0.2 0.1 1.0 — 0.7 0.4 8.8
B1-8(1) — — — n.a. 0.2 1.0 0.1 0.4 0.3 0.2 0.3 0.2 2.8
B1-10 — — — n.a. 0.3 1.3 0.2 0.3 0.6 1.9 0.8 0.3 5.6
B1-12 — — — 0.3 0.8 1.1 0.2 0.2 0.9 1.8 0.7 0.3 6.3
B1-14 — — — n.a. 0.1 0.9 0.1 0.3 0.4 0.7 1.0 0.2 3.7
B1-8(2) — — — n.a. 0.2 1.2 0.2 0.3 0.3 1.0 n.a. 0.4 3.7
Concentration, µg/sampleSample Sum,
µg/sampleLocation Depth, ft Temp, ºF
*
* * *
* Wick adhered to silicon pressure pad, and/or reservoir pad was only partially saturated
17
Summary • Dusts on Calvert Cliffs and Hope Creek canisters are largely insoluble
minerals; salts are limited, and are salts are largely Ca-sulfate and nitrate-rich. NaCl was observed as rare isolated grains.
• Dusts on Diablo Canyon canisters are sea-salt rich. Sea-salts are present are present in both the fine (<2.5µm) and coarse (>2.5µm fraction). Larger grains are spherical aggregates or euhedral crystals of halite, with associated Mg-sulfate, and lesser amounts of Ca and K.
Field data indicate that in at least some near-marine ISFSI locations, chloride-rich sea-salt aerosols comprise a large fraction of dusts deposited on canister surfaces. Once deliquescence occurs, SCC may be possible.
18
Implications
Experimental data suggest that SCC will occur when sea-salts deliquesce. But, are experimental conditions typical of field conditions?
Aqueous solution deposited by airbrush:
Salts in ethanol deposited by airbrush. Salts on waste package
surface:
2 µm
2 µm
2 µm
4 µm 250 µm
120 µg/cm2
18
Extra slides
19
Samples Collected — Hope Creek
20
MPC-145 MPC-144
Wick stuck to silicone pressure pad, and/or reservoir pad was only partially wetted
0
2
4
6
8
10
12
14
1620 30 40 50 60 70
Dept
h of
inse
rtio
n (ft
)
Temp. (ºC)
MPC-144
Wet sample
Dry sample
Temp. meas.
0
2
4
6
8
10
12
14
160 25 50 75 100
Dept
h of
inse
rtio
n (in
)
Temp. (ºC)
MPC-145
Wet sample
Dry sample
Temp. meas.
Samples Collected — Diablo Canyon
21
MPC-170 MPC-123
Wick stuck to silicone pressure pad, and/or reservoir pad was only partially wetted
0
2
4
6
8
10
12
14
1640 60 80 100 120 140
Dept
h of
inse
rtio
n (ft
)
Temp. (ºC)
MPC-123
Wet sample
Dry sample
0
2
4
6
8
10
12
14
1660 70 80 90 100
Dept
h of
inse
rtio
n (ft
)
Temp. (ºC)
MPC-170
Wet sampleDry sampleTemp. meas.
Sample photographs
22
144-003 144-004 144-008
144-009 144-010 144-013
144-014
144-001 144-002
144-005 144-006
144-007 144-011
144-012
Samples collected from Hope Creek MPC-144
Photos placed in horizontal position with even amount of white space
between photos and header
Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-AC04-94AL85000. SAND2014-18436 PE
Current DOE Used Fuel Disposition Storage and Transportation R&D Activities
Sylvia Saltzstein, Sandia National Labs BAM and SNL Collaboration Workshop
October 6-9, 2014, Albuquerque, NM, USA
Scenario Difficulty
Con
seq
uenc
e
Contents Overall Storage and Transportation R&D Objectives DOE High Burnup Dry Storage Cask R&D Project
Status of High Burn-up related R&D work in technical
Control Accounts Field Demonstration Experiments Transportation Analysis Security
2 www.nrc.gov/waste/spent-fuel-storage/ www.connyankee.com/
http://energy.gov/sites/prod/files/styles/
3
Overall Objectives: 1. Develop the technical bases to demonstrate high burn-up used fuel
integrity for extended storage periods. 2. Develop technical bases for fuel retrievability and transportation after
long term storage. 3. Develop the technical basis for transportation of high burnup fuel.
Storage and Transportation R&D Objectives
Storage System Component “High” and “Medium” priorities
System Component Issue Importance of R&D
Cladding
Annealing of Radiation Effects Medium Oxidation Medium H2 effects: Embrittlement High H2 effects: Delayed Hydride Cracking High
Creep Medium Assembly Hardware Stress corrosion cracking Medium
Neutron Poisons
Thermal aging effects Medium Embrittlement and cracking Medium Creep Medium Corrosion (blistering) Medium
Canister
Atmospheric corrosion (marine environment)
High
Aqueous corrosion
High Source: Gap Analysis to Support Extended Storage of Used Nuclear Fuel, January 2012
Storage System Component “High” and “Medium” priorities
System Component Issue Importance of R&D
Bolted Direct Load Casks
Thermo-mechanical fatigue of bolts/seals Medium
Atmospheric corrosion (marine environment) High
Aqueous corrosion High
Overpack and Pad (Concrete) Freeze/Thaw Medium Corrosion of steel rebar Medium
Cross-cutting or General Gaps
• Temperature profiles for fuel High • Drying issues High • Monitoring High • Subcriticality High • Fuel transfer options High • Re-examine INL dry cask storage High
Identification of these data gaps are used to inform new initiatives for FY15
Source: Gap Analysis to Support Extended Storage of Used Nuclear Fuel, January 2012
FULL-SCALE HIGH-BURNUP DEMO Purpose: To collect data on high-burnup fuel in realistic storage conditions.
High Burnup Dry Storage Confirmatory Demo: Goals
Goals: provide confirmatory data for
model validation and potential improvement,
provide input to future SNF dry storage cask design,
support license renewals and new licenses for Independent Spent Fuel Storage Installations (ISFSIs), and
support transportation licensing for high burnup SNF.
View of Fuel Basket within a Typical TN-32 Cask
Lid Installation of a TN-32 Cask
High Burn-up Confirmatory Data Project: Timeline
Picture from North Anna ISFSI
2017: Load a TN-32B storage cask with high burn-up fuel in a utility storage pool • Loaded with well-understood fuel •Remove sister pins for baseline analysis and data collection (some sister pins will have been pulled in 2015) • Cask will have instrumentation for monitoring
2017: Dry the cask contents using typical process
2017- 2028: House cask at the utility’s dry cask storage site (North Anna) •Continually monitored and inspected for >10 years.
2028: Open cask investigate condition of fuel. (Location TBD)
High Burn-up Confirmatory Data Project: Data to be Monitored Fuel cladding temperature (indirect via thermocouple lances)
Cavity gas monitoring is being evaluated to check for damaged
fuel and residual water Pressure Composition
Fission gasses Moisture Hydrogen Oxygen
Active methods for sampling the gas were analyzed Use of remote sensors were evaluated to gather the needed data Gas sampling on the pad is still be investigated
Picture from North Anna ISFSI
High Burn-up Confirmatory Data Project: Rod Testing to Establish the Baseline
Picture from North Anna ISFSI
Testing of similar rods as those to be loaded in the cask Some fuel rods (25 or less) will be shipped in existing licensed cask to
a hot cell for baseline rod characteristic data Location to receive the shipment is still under discussion
Some rods will come from sister assemblies and some rods from assemblies to be stored in the TN-32
Schedule for obtaining pins of similar nature as to be loaded
in the cask (similar pins) Similar pins will be pulled in 2015 Similar pins will be shipped in 2015 or 2016
11
High Burn-up Confirmatory Data Project: On-going Sensor Technology Development Assess sensor technologies to interrogate future
dry storage canister systems for: crack characteristics associated with stress
corrosion cracking thermal conditions humidity conditions fission gas release
Collaborate with industry to align sensor technologies with operational constraints
Support dry storage license extension certification efforts
Support confidence in licensee’s ability to detect cracks (1st), assess crack size (2nd), crack volume (3rd). 5 year proposed inspection interval.
EXPERIMENTS Purpose: Collect data on material properties, and environmental conditions that could affect performance.
Experiments: High Burnup Fuel Cladding Material Properties
Separate effects test to determine effects of hydrides, hydride reorientation, radiation damage, thermal annealing, and clad thinning on materials properties and performance.
Hydrides and reorientation Ring Compression Tests and determination
of Ductile-Brittle Transition Temperature (ANL)
Cladding bend test and effects of fuel/clad bonding and pellet/pellet interfaces (ORNL)
Radiation damage and thermal annealing Irradiate cladding in HFIR reactor at ORNL
without all other effects.
Jy-An, Wang; Oak Ridge National Laboratory, WM2014 Conference, March 2014
Billone, Argonne National Laboratory, EPRI ESCP Meeting, Dec. 2013
DBTT data for Ziro clad with Varying Internal Plenum Temperatures
Used fuel rod stiffness Experiments (in hot cell and out) and analyses of
stress distribution
Experiments: Stainless Steel Canister Corrosion Purpose: Better understand canister
degradation, support Aging Management Plans, and license extensions.
Develop data to understand initiating conditions for corrosion conditions and progression of SCC-induced crack growth
Obtain site data to assess atmospheric conditions and compare with initiating conditions.
Procure a full scale (diameter) welded SS canister to investigate residual stresses due to plate rolling and welding.
Enos, et al., Data Report on Corrosion Testing of Stainless Steel SNL Storage Canisters, FCRD-UFD-2013-000324
Dust on top surface of SS canister
Dust particles on filter
Conceptual design for full-scale (diameter) SS welded canister
Collecting dust samples at Diablo Canyon
Sea Salt crystal with MgSO4 inside found on Diablo Canyon Canister
TRANSPORTATION Purpose: Will the fuel remain intact during transportation?
Transportation: Normal Conditions of Transport – Loading on fuel assemblies
16
A surrogate assembly was subjected to truck data from a 700 mile trip on a shaker table and 50 miles on a real truck with representative weight. Data results were >10 times below yield
strength. The strains measured in both were an
order of magnitude lower than either an irradiated or unirradiated Zircaloy rod yield strength.
If high burnup fuel can maintain its integrity during transport, pressure will be taken off experimental R&D efforts associated with hydride effects on cladding strength and ductility.
200 μϵ measured 700 μϵ computed
Sorenson, K., Determination of Loadings on Spent Fuel Assemblies During Normal Conditions of Transport,
SAND2014-2043P.
7000 - 9000 μϵ @ yield
Data collection and analysis for NCT loads on a surrogate fuel assembly
ANALYSIS Purpose: Develop predictive models of material behavior to establish the technical bases for extended storage and transportation.
Analysis
18
Predictive modeling Thermal Analysis (PNNL) to predict cool
down, Ductile to Brittle Transition, deliquescence, etc.
HBU Demonstration fuel selection and cool down
Modern, high heat load, high capacity systems
In-service inspections validation data
Hybrid hydride reorientation model (SNL)
Structural uncertainty analysis at assembly and canister level (PNNL)
Finite element analysis validation with CIRFT and application to out-of-cell testing (ORNL)
Thermal profile analyses Detailed thermal analyses for 2-3 licensed
dry storage systems (PNNL FY15)
CFD Thermal Analysis of Dry Storage Casks
Suffield, et al, PNNL-21788
Model for Simulation of Hydride Precipitation, Tikare et
al, FCRD-UFD-2013-000251.
FE Models of Assembly Klymyshyn, et al, PNNL, FCRD-UFD-
2013-000168
Jy-An, Wang; Oak Ridge National Laboratory, WM2014 Conference, March 2014
SECURITY Purpose: Understand our vulnerabilities and how to mitigate risk.
20
The RIMES methodology focuses on the degree of difficulty for an adversary to successfully accomplish an attack
An expert panel will be used to develop scenarios and determine the degree of difficulty This work builds off the MPACT work on used fuel storage security.
Scenario Difficulty
Con
sequ
ence
Attack scenarios that are both easier and high consequence are of greater risk. Focus security investments on these “high-risk” scenarios.
Security: Assessing Transportation Security Risk
STRATEGIC INITIATIVES Purpose: What are the most important things for us to do?
1. Reviewed and summarized all (>180) DOE UFD reports written from 2010 to 2014.
2. Categorized UFD Reports into 15 high and medium
gaps from previous Gap Analyses. Hydride reorientation and embrittlement, Welded canister-atmospheric corrosion, Bolted casks-embrittlement of seals, Drying…
3. Summarized for each gap: 1. What we have learned 2. What we still need to learn 3. Revised ranking 4. Determination to continue or defer R&D efforts
during the next three years.
22
UNF Extended S&T R&D Review and Plan
Purpose: To develop a methodology that will identify what data is the most important to close the technical gaps. Identify performance characteristics of a
degradation mechanism. CISCC
Link the degradation mechanisms to the regulatory requirements. Ex. no through-wall crack penetration.
Understand the currently available data and identify the uncertainties with that data.
Perform decision making analysis Identify areas with insufficient data and rank.
Final product should be a prioritized list of the most and least impactful data to close the gaps.
Uncertainty Quantification
23
THANK YOU! QUESTIONS?
SNL&BAM S&T R&D Collaborative Workshop
October 6 – 8, 2014 IPB/1150
Agenda
Monday, October 6, 2014
18:30 – 20:00 Welcome and BBQ followed by general discussion Whole Hog Café 9880 Montgomery Blvd NE,
Tuesday, October 7, 2014 10:00 – 10:15 SNL Introduction Tito Bonano IPB/1150 10:15 – 10:30 BAM Introduction Bernhard Droste IPB/1150 10:30 – 11:00 Overview of S&T at SNL Sylvia Saltzstein IPB/1150 11:00 – 11:30 Overview of DR at SNL Kevin McMahon IPB/1150
11:30 – 12:00 Extended interim storage issues and long-term investigations at BAM Holger Völzke IPB/1150 12:00 - 13:00 Lunch Dion’s 13:00 – 13:30 Travel to Brayton Cycle Building 13:30 – 14:15 Tour of Brayton Cycle Darryn Fleming 6630 14:30– 15:00 Tour of CyBl G. Koenig, E. Lindgrin 6922/6536 15:00 – 15:30 Travel back to IPB from CyBl 15:30 – 16:00 Truck Transport Results/Progress on Rail Test Paul McConnell IPB/1150 16:00 – 16:30 Transportation Risk Communication Doug Ammerman IPB/1150 16:30 – 17:00 Transportation Logistics Elena Kalinina IPB/1150
Wednesday, October 8, 2014 08:30 – 09:00 Design Leakage Rates for Activity Release Calculation Annette Rolle IPB/1150 09:00 – 9:30 SNF/HLW dual and multi-purpose cask issues Bernhard Droste IPB/1150 9:30 – 10:00 DPCs Tito Bonano IPB/1150 10:00 – 10:15 Break 10:15 – 10:45 Cooperation BAM/ITU on Hot Cell Testing of SNF Rod Segments Konrad Linnemann IPB/1150 10:45– 11:15 SCC and Full Scale Weld Update Charles Bryan, David Enos IPB/1150 11:15 – 11:30 Reflections on Morning Topics 11:30 – 13:00 Lunch 13:00 – 13:30 Knowledge Management Kevin McMahon IPB/1150 13:30 – 14:00 Integrating Mgmt of SNF from Generation to Disposal R. Rechard, L. Price, E. Kalinina 14:00 – 14:30 Social Science Update Hank Jenkins-Smith, IPB/1150 Kuhika Gupta-Ripberger , R. Rechard
14:30 – 17:00 Strategic Path Forward Sylvia Saltzstein/All IPB/1150
17:00 Adjourn SAND2014-19330 O
Holger Völzke
SNL - BAM - Workshop
October 06-08, 2014, Albuquerque, NM,USA
1
Extended Interim Storage Issues and Long-Term Investigations at BAM
Holger Völzke
BAM Bundesanstalt für Materialforschung und –prüfung
(Federal Institute for Materials Research and Testing)
Berlin, Germany
Outline:
1. Present Status of SNF and HLW Management in Germany
2. Operation Experience and Regulatory Framework in DPC Storage
3. Perspectives and Challenges Concerning Extended Interim Storage
4. Current Long-term Investigations at BAM
5. Conclusions
Holger Völzke
SNL - BAM - Workshop
October 06-08, 2014, Albuquerque, NM,USA
2
Until 2011:
17 NPPs with 21,5 GWe
Since Aug. 2011 (After phase out decision and Atomic Energy Act revision):
9 NPPs with 12,7 GWe remaining
Further reactor shut downs until 31.12.2022 !
1. Present Status of SNF and HLW Management in Germany
Accumulated spent fuel until 31.12.2013:
≈ 14,900 Mg HM (≈ 53,600 fuel assemblies)
Estimated amounts until final reactor shut-down (incl. cores):
≈ 2,300 Mg HM (≈ 6,400 fuel assemblies)
Total amount until 31.12.2022:
≈ 60,000 fuel assemblies
≈ 17,200 Mg HM
At present ≈ 1,000 dual purpose casks of various types are in use
for dry storage of SNF and HAW - at 16 storage sites (12 on-site)
About 500 …600 additional casks needed during the next decade.
About 50 … 80 casks every year to be manufactured, loaded and stored
Additional transport and storage licenses needed for various fuel data
(e.g. burn-up) and also defect fuel assemblies
NPP Grafenrheinfeld
Holger Völzke
3
SNL - BAM - Workshop
October 06-08, 2014, Albuquerque, NM,USA
Cumulated Generated Spent Fuel
6,700 Mg
HLW from
reprocessing (SF delivery ended
by 30.06.2005)
Final reactor
shut-down 2022
10,500 Mg
Spent Fuel
Today
1. Present Status of SNF and HLW Management in Germany
Holger Völzke
4
SNL - BAM - Workshop
October 06-08, 2014, Albuquerque, NM,USA
Central SF Storage Facilities (since 1982)
12 On-site SF Storage Facilities (since 2002-2005)
Source: BfS
Interim Storage North near Greifswald with
CASTOR 440/84 casks containing VVER fuel
(Photo: EWN GmbH)
Jülich Research Center: 20 year storage license for AVR fuel expired
June 30, 2013 !
Interim Storage North
for VVER Fuel
1. Present Status of SNF and HLW Management in Germany
Holger Völzke
5
SNL - BAM - Workshop
October 06-08, 2014, Albuquerque, NM,USA
12 On-Site Storage Facilities – 3 Designs
Tunnel Design (NPP Neckarwestheim)
Steag Design
WTI Design
1. Present Status of SNF and HLW Management in Germany
Interim Storage Facility at Isar NPP
Holger Völzke
SNL - BAM - Workshop
October 06-08, 2014, Albuquerque, NM,USA
6
Accident safe dual purpose transport and storage casks
Valid Type B(U) approval required before loading and
during storage to guarantee permanent transportability
Monolithic thick walled metal cask body
Vacuum dried and helium filled (≈ 800 hPa) cask interior
inert conditions
Permanently monitored double barrier lid system
equipped with metal seals
Qualified repair concept in case of hypothetical lid failure
Casks stored inside buildings
Current storage licenses limited to 40 years
German Dry Spent Fuel and HLW Storage Concept
Photos: GNS
1. Present Status of SNF and HLW Management in Germany
Holger Völzke
7
SNL - BAM - Workshop
October 06-08, 2014, Albuquerque, NM,USA
2. Operation Experience and Regulatory Framework in DPC Storage
Storage License
issued by BfS (Federal Office for Radation Protection)
Licensee/Operator
Supervising State Authority
Contracted Experts Contracted Experts
Licensing Procedure Operational Regime
Storage operation since:
• 1992 CASTOR® THTR/AVR (TBL Ahaus)
• 1993 CASTOR® THTR/AVR (Jülich)
• 1997 CASTOR® Ic, V/19, HAW 20/28CG (TBL Gorleben)
• ….
> 20 years of safe storage operation without
safety relevant issues (e.g. no seal failure)
Holger Völzke
8
SNL - BAM - Workshop
October 06-08, 2014, Albuquerque, NM,USA
2. Operation Experience and Regulatory Framework in DPC Storage
§6 of Atomic Energy Act (AtG) - Act on the Peaceful
Utilization of Atomic Energy and the Protection against its
Hazards
“Guidelines for Dry Cask Storage of Spent Fuel and Heat-generating Waste” German Waste management Commission (ESK), Revised version of 10.06.2013, http://www.entsorgungskommission.de/englisch/downloads/eskempfehlungesk30llberevfassung10062013en.pdf
“ESK-Guidelines for Periodic Safety Inspections and Technical Ageing
Management for Interim Storage Facilities for Spent Nuclear Fuel and Heat-
generating Radioactive Waste” German Waste management Commission (ESK), Version of 13.03.2014, http://www.entsorgungskommission.de/downloads/empfehlungpsuzl13032014homepage.pdf
Nuclear Waste Management Commission
The Storage License contains all relevant safety evaluations to satisfy
the protection goals (safe enclosure, shielding, subcriticality, heat
dissipation) under operational and accidental conditions of the specific
storage facility and defines conditions and requirements for its safe and
secure operation.
Holger Völzke
9
SNL - BAM - Workshop
October 06-08, 2014, Albuquerque, NM,USA
2. Operation Experience and Regulatory Framework in DPC Storage
Type B(U)
package design approval
+
Safety inspections prior to
transportation
Holistic approach merging
requirements from storage
licenses and type B(U) approvals
Storage License
incl. specific cask types
and inventory
Continuous aging
management
Periodic safety reviews
every 10 years
Permanent
transportability of all
stored casks
Ruled by the German Atomic Energy Act
Ruled by the Transport Regulations for
Dangerous Goods under consideration
of IAEA Regulations for the
Transportation of Radioactive Materials
Timeframe for aging
considerations:
40 years? Longer periods? Aging management has to
consider transport and storage
needs and has to be inserted in
the operational storage regime
Holger Völzke
10
SNL - BAM - Workshop
October 06-08, 2014, Albuquerque, NM,USA
3. Perspectives and Challenges Concerning Extended Interim Storage
Expiration of present interim storage licenses:
On-site Storages 2042/43
Interim Storage North 31.10.2039
Gorleben 31.12.2034
Ahaus 31.12.2036 (first CASTOR® THTR/AVR cask already 2032 !)
Jülich 30.06.2013 ! (After only 20 years !)
2016
Site selection
procedure
2023
Site selection for
underground
exploration
2031
Final site
selection
Licensing procedure
and repository
construction 15 … 20 years
≈ 2050 Operation phase
30 years
(50 casks per year)
2032/2036 2042/2043
Expiration of
interim storage
licenses
At least 40 years
extended interim
storage
Timeline without any delays caused by lawsuits !!!
Timeline derived from the current Site Selection Law for Disposal:
≈ 2080
Holger Völzke
11
SNL - BAM - Workshop
October 06-08, 2014, Albuquerque, NM,USA
3. Perspectives and Challenges Concerning Extended Interim Storage
Storage License
Continuous aging
management
Periodic safety reviews,
every 10 years
Extended
Storage
Additional safety assessments
concerning
degradation effects
Transportation
after Storage
Transportation
after Storage
Type B(U) Approval
+
Safety inspections prior to
transportation
Path Forward to Extended Storage
Basic R&D programs
Final Disposal
Holger Völzke
12
SNL - BAM - Workshop
October 06-08, 2014, Albuquerque, NM,USA
3. Perspectives and Challenges Concerning Extended Interim Storage
Dual Purpose Casks
During Interim Storage Initial cask loading
and storage
Initial package
design approval
for transportation
Cask operation:
- Storage
- Handling
- Maintenance
- Aging Management
Initial storage
period: 40 years
Extended storage
period(s):
20, 40, 60, … (?) years
Validity up to
10 years
Design approval
prolongation
Periodic safety
inspections
every 10 years
Requirements / Outcomes
Holger Völzke
SNL - BAM - Workshop
October 06-08, 2014, Albuquerque, NM,USA
13
Outcomes from operation experience, ageing management programs, and periodic
safety inspections
Development of the technical and scientific state-of-the-art with regard to necessary
precautions against damages by the storage of nuclear fuel
Changes of regulatory requirements
Additional safety assessments concerning degradation effects
for extended storage periods and transportation after storage
3. Perspectives and Challenges Concerning Extended Interim Storage
Consideration of existing casks and their inventories
Data needs concerning storage periods beyond 40 years:
Long-term performance of bolted lid systems
• Bolt relaxation
• Metal seal relaxation and creep
• Other material degradation by temperature, time, ambient conditions
• Leakage rate measurements after long storage periods concerning elastomer auxiliary
seal degradation and helium contamination
• Reliability of pressure monitoring devices
Degradation of polymer components for neutron shielding
Safety margins of aged casks in severe accident scenarios
Long-term performance of cask inventories (fuel assemblies, canisters, baskets)
Photo: GNS Resulting leak-tightness
Holger Völzke
SNL - BAM - Workshop
October 06-08, 2014, Albuquerque, NM,USA
14
4. Current Long-term Investigations at BAM
Test Parameters Temperatures: +20°C +75°C +100°C +125°C +150°C
Holding times since: 02/2009 01/2014 11/2010 01/2014 02/2009
Seal type Al + Ag Al + Ag Al + Ag Al + Ag Al + Ag
BAM laboratory tests with continuous
leakage rate measurement during seal
loading and unloading
Helicoflex®
metal seals
Long-term performance of Helicoflex® metal seals
0
200
400
600
0.00 0.50 1.00 1.50
Load initial compression
Relieving after 1 week
Rel. after 5 weeks
Rel. after 3 months
Rel. after 6 months
Rel. after 12 months
Rel. after 18 months
Rel. after 25 months
Rel. after 33 months
Rel. after 36 months
Rel. after 44 months
Rel. after 48 months
Deformation [mm]
Lo
ad
[N
/mm
]
Reduction
of restoring
seal forceFr
Reduction of ru
Y2
Y1
Al-seal 150 C
Holger Völzke
SNL - BAM - Workshop
October 06-08, 2014, Albuquerque, NM,USA
15
4. Current Long-term Investigations at BAM
Restoring seal force Fr (Load) reduction
depending on holding time and temperature
- for test periods up to 48 months and
- extrapolation up to 100 years (dashed lines)
0
200
400
600
0.5 5 50 500 5000
Lo
ad
[N
/mm
]
Holding Time [weeks]
20 C
100 C
150 C
Al-Seals
40 Years
ca. 290 N/mm
ca. 220 N/mm
ca. 85 N/mm
100 Years
77%
31%
0
200
400
600
0.5 5 50 500 5000
Lo
ad
[N
/mm
]
Holding Time [weeks]
150 C
20 C
100 C
Ag-Seals
40 Years
ca. 400 N/mm
ca. 280 N/mm
ca. 160 N/mm
100 Years
80%
42%
Metal seal test results (1):
Ref.: Holger Völzke et al., Paper #104, Proceedings of the 17th International Symposium on the
Packaging and Transportation of Radioactive Materials PATRAM 2013, August 18-23, 2013, San Francisco, CA, USA
Holger Völzke
SNL - BAM - Workshop
October 06-08, 2014, Albuquerque, NM,USA
16
4. Current Long-term Investigations at BAM
(U)HMW-PE for neutron radiation shielding
CASTOR® HAW28M
cask design by GNS
Requirement:
Sufficient long-term neutron radiation shielding
without safety relevant degradation
Degradation effects:
Temperatures (max. 160°C; decreasing during storage) Thermal expansion
Structural changes from semi-crystalline to amorphous
Gamma radiation (decreasing during storage) Structural damages and/or crosslinking
hydrogen separation
Mechanical assembling stresses Stress relaxation
Gamma irradiation tests by BAM (at room temperature)
• Low dose irradiation (60Co source): 0.5 – 60 kGy
• High dose irradiation (conservative max. storage dose): 600 kGy
Holger Völzke
SNL - BAM - Workshop
October 06-08, 2014, Albuquerque, NM,USA
17
4. Current Long-term Investigations at BAM
Further gamma irradiation tests with
material blocks of 10*10*50 cm³
0 kGy
50 kGy
100 kGy
200 kGy
400 kGy
600 kGy
UHMW-PE HMW-PE
Outcomes for UHMW-PE from various analyses (exemplary): Increase of insoluble, crosslinked fraction after high dose gamma irradiation
Future investigations planned:
Thermal aging of(U)HMW-PE at elevated
temperatures
Combination of radiation and thermal aging
Development of adequate prognostic
methods to allow extrapolation of long-term
material performance
Holger Völzke
SNL - BAM - Workshop
October 06-08, 2014, Albuquerque, NM,USA
18
4. Current Long-term Investigations at BAM
Investigation of Elastomer Seals
Used as auxiliary seals in spent fuel and HLW casks
Used as primary seals in LLW/ILW casks
Points of interest:
1. Low temperature behavior down to -40°C
Recent Publications by Matthias Jaunich, Wolfgang Stark, and Dietmar Wolff:
Low Temperature Properties of Rubber Seals
Kgk-Kautschuk Gummi Kunststoffe, 2011. 64(3): p. 52-55.
A new method to evaluate the low temperature function of rubber sealing materials
Polymer Testing, 2010. 29(7): p. 815-823.
Comparison of low temperature properties of different elastomer materials
investigated by a new method for compression set measurement.
Polymer Testing, 2012. 31(8): p. 987-992.
Holger Völzke
SNL - BAM - Workshop
October 06-08, 2014, Albuquerque, NM,USA
19
4. Current Long-term Investigations at BAM
2. Aging under themo-mechanical loads (and by irradiation)
Investigation program with selected rubbers (HNBR, EPDM and FKM) tested as
O-rings with an inner diameter of 190 mm and an cross sectional diameter of 10
mm has been started in May 2014.
The O-rings are oven-aged at four different temperatures (75 °C, 100 °C, 125 °C, 150 °C).
They are examined after various times (1 d, 3 d, 10 d, 30 d, 100 d, 0.5 a, 1 a, 1.5 a, 2 a, 2.5 a, 3 a,
3.67 a, 4.33 a and 5 a).
In order to be able to compare between compressed and relaxed rubber, the
samples are aged in their initial O-ring state (Fig. 1) as well as compressed
between plates (Fig. 2) with a deformation of 25 % corresponding to the actual
compression during service. Furthermore, we are aging samples in flanges that
allow leakage rate measurements (Fig. 3).
Fig. 1 Undeformed O-rings
Fig. 2 Half O-rings compressed between plates Fig. 3 O-ring in flange for
leakage measurements
Holger Völzke
SNL - BAM - Workshop
October 06-08, 2014, Albuquerque, NM,USA
20
4. Current Long-term Investigations at BAM
Present BAM R&D Perspectives
Investigation of Helicoflex metal seals (Ag- and Al-seals) concerning the
long-term behavior at various temperatures between 20°C and 150°C
Main objectives: Determination of • Restoring seal force (Fr) reduction
• Useable resilience (ru) reduction
Outcomes from 48 months test period • Plasticization of the outer seal jacket
• Reduction of Fr and ru during long-term loading
• Increased seal function due to improving material contact
• Linear logarithmic correlation and extrapolation of Fr and ru up to 100 years
Further plans • Continuation of running seal tests towards longer periods of time
• Investigation of additional temperature levels
• Evaluation of the Larson-Miller approach
Investigation of polymers for neutron shielding • Material property changes due to mechanical and
thermal aging and irradiation
Investigation of elastomer seals • Low temperature behavior
• Aging by mechanical and thermal loads and irradiation
Holger Völzke
SNL - BAM - Workshop
October 06-08, 2014, Albuquerque, NM,USA
21
For more than 20 years dry interim storage of spent fuel and HLW in approved
dual purpose casks has demonstrated safe, secure, reliable, and flexible
operation without any failures.
Germany’s final phase-out decision 2011 and the complete restart of the
repository siting procedure in 2013 result in new challenges for near-term cask
availability and provoke the need for extended interim storage periods
Extended storage periods of 80 years or even more are most likely
Storage and transportation (after storage) are closely linked
Further improvement of the regulatory framework gives valuable support
Ageing management issues should be addressed in a holistic approach
Additional safety demonstrations and material data for the long-term concerning
casks and inventory will be required in the future and should consider both
storage and transportation needs
First R&D initiatives have already been started by BAM concentrating on metal
seals, polymers for neutron shielding, and elastomer seals
5. Conclusions
Bernhard Droste 1
A Survey on the History of Relations between BAM and Sandia NL
Bernhard DrosteBAM Federal Institute for Materials Research and Testing
Berlin, Germany bernhard.droste@bam.de
BAM/Sandia WorkshopAlbuquerque, NM, USA, October 6-8, 2014
Bernhard Droste BAM/Sandia Workshop 2
High speed impact testing of a modified
18/8 Container for Pu-nitrate
Sandia test site May 15, 1979
BAM/Sandia Cooperation after meetings at PATRAM 1978, Las Vegas
Bernhard Droste BAM/Sandia Workshop 3
BAM/Sandia Cooperation 1979
Bernhard Droste BAM/Sandia Workshop 4
Joint Sandia/BAM Research on Transport Package Safety 1985-1992
Bernhard Droste BAM/Sandia Workshop
Hallo
Joint BAM/Sandia Drop Test Program
14m drop test with the CASTOR VHLW cask (with a
120 mm deep failure inside the 260 mm ductile iron
wall) in comparison with the regulatory 9-m drop test
Regulatory tests for US approval accompanied by Sandia measurements at BAM test site Lehre
5
Joint Sandia/DOE/BAM Research on Safety of Ductile Iron Casks 1985-1992
14 m drop onto steel rolls,
without impact limiter
9 m drop onto unyielding target,
equipped with impact limiter
Time [ms]
De
ce
lera
tion [g]
Bernhard Droste BAM/Sandia Workshop
Research in Safety of Ductile Iron Casks 1987
BAM drop test with a thick-walled pipe of ductile cast iron
• corresponding to the 1:2.5 scaled model of a large cylindrical CASTOR V cask
• drop height 9 m
• drop onto steel cylinders located on an unyielding IAEA target
• equipped with an artificial crack-like defect (40 mm in 150 mm wall)
6
Bernhard Droste7
BAM/Sandia Workshop
SANDIA National Labs. Drop Tests with MOSAIK Cask
Sequenz von Fallversuchen mit MOSAIK KfK mit künstlichem
rissartigem Fehler auf ein Stahlrollenlager auf IAEA-Fundament
bei -29 °C
Fehlstelle nach dem 5. Fallversuch:
< 1 mm duktile Rissverlängerung
Versuch 1 bis 4: 9 m Fallhöhe, Risstiefe bis 76 mm (36% Wanddicke) � keine Rissinitiierung
Versuch 5: 18 m Fallhöhe, Risstiefe 57 mm � Risswachstum ohne sprödes Versagen
Joint Sandia/DOE/BAM Research on Safety of Ductile Iron Casks 1985-1992
Bernhard Droste BAM/Sandia Workshop 8
Signing of the project agreement DOE/BAM
Paris, PATRAM`98
(Bernhard Droste, Kelvin Kelkenberg)
DOE/BAM Agreement 1998, RadWaste Transport
Bernhard Droste BAM/Sandia Workshop 9
DOE/BAM/SNL Workshop, Albuquerque, 1999
Bernhard Droste BAM/Sandia Workshop 10
DOE/BAM/SNL Workshop, Albuquerque, 1999
Technical tour to WIPP, Carlsbad
…in front of the first badges of
TRU waste inside the repository
(Florentin Lange/GRS, Mona Williams/DOE, Bernhard Droste/BAM, Ashok Kapoor/DOE,Richard Yoshimura/SNL, Uwe Zencker/BAM)
Bernhard Droste
Visit of US National Academies WG
for Spent Fuel Cask Full-Scale Drop Testing
BAM TTS, September 24, 2004
(Technical Tour 2, PATRAM 2004)
Full-Scale Model MSF 69 BG (MHI)
• Total Mass: 127,000 kg
• Length with Impact Limiters: 6,800 mm
• External Diameter of Impact Limiters: 3,100 mm
BAM/Sandia Workshop
US National Academies` Visit 2004
11
Dr. Bernhard Droste 12
Cooperative Agreement
BAM/U.S. Nuclear
Regulatory Commission
US NRC PosterPATRAM 2010, London, on comparison of NRC calculations with BAM measurements of a 9-m drop test with the CONSTOR V/TC full scale cask (GNS)
BAM`s intention was also to come to closer cooperation with Sandia NL in the proposed Package Performance Study (PPS)….which was cancelled later on.
BAM/US NRC Cooperation 2006-2010
Bernhard Droste BAM/Sandia Workshop 13
BAM/US NRC Cooperation
Visit of US NRC Commissioner Ms K.L.Svinicki
in Berlin and at BAM Test Site, March 24, 2010
Bernhard Droste BAM/Sandia Workshop 14
Conclusions, Recommendation:
Proceed with Cooperation!!!!
Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-AC04-94AL85000.
Integrating Management of Spent Nuclear Fuel from Generation to
Disposal in the United States
Rob P. Rechard Laura L. Price Elena Kalinina
Sandia National Laboratories
Storage and Transportation Workshop between Sandia and BAM Albuquerque, New Mexico
7-8 October 2014 SAND2014-18539 PE
2
US History of Commercial Power Reactors
• 9 Early Prototypes – No fuel on site
• 1 Never Operated
• 1 Disabled (Three Mile Island) – Fuel moved to DOE
• 1 Demonstration High Temperature Gas Reactor (Fort St. Vrain in Colorado)
• 18 Ceased Operations – Fuel on site
– 3 reactors on sites with on going nuclear operations
– 15 reactors on 12 sites with no other nuclear operations
• 100 Operating Reactors
• 6 New Reactors at Existing Sites Under Construction
130 Commercial Nuclear Power Plants Built
World Institute for Nuclear Security, June 10-12, 2014 Jeff Williams
3
Current waste management system uses at-reactor storage
•100 operating reactor at 62 sites in 2014
– 65 pressurized water reactors (PWR)
– 35 boiling water reactors (BWR)
•71,000 tonnes heavy metal radioactive waste in 2013
– 49,000 tonnes in wet storage
– 22,000 tonnes in dry storage
IHLRWMC
4/30/2013
4
Licensing of storage is deterministic and rule-based in US
•Wet storage licensed as part of reactor operations – Reactor license for up to 60 y, with 20 y renewal – 10 CFR 50
•Dry storage licensed separately – 69 Independent Spent Fuel Storage Installations
(ISFSI) in 2013 – Licensed up to 40 y with up to 40 y extensions – 10 CFR 72
•2 types of ISFSI licenses – 54 General licenses
- Co-located with operating reactor - 3.5 y to complete application
– 15 Site-specific licenses - Separate from reactor or reactor is shut down - 6 y to complete application IHLRWMC
4/30/2013
5
Several types of ISFSI designs in US
•Vertical above ground •Vertical below ground •Horizontal bunker •1 Vault: DOE site in Colorado for Fort St.
Vrain SNF (high temperature gas cooled reactor)
6
Dry74,197, 30%22,000 MT1,850 Casks
Pool172,281, 70%49,024 MT
Transnuclear TN-32 Holtec Hi-Star 100
Dry Storage Inventory
Majority is in Large Welded Canisters
Current dry storage inventory is diverse
Trend toward higher capacities
Transnuclear (34%) Holtec (41%)
NAC (10%)
1,655 Welded Metal Canisters In Vented Concrete Overpacks 65,102 Assemblies, 87.5% of Dry
183 Bare Fuel Casks 8,406 Assemblies, 11.3% of Dry
12 Welded Metal Canisters in Transport Overpacks 866 Assemblies, 1.2% of Dry
World Institute for Nuclear Security, June 10-12, 2014
7
Shutdown Reactor Sites Use Several Different Storage Designs
Humboldt Bay, Holtec below grade
Maine Yankee, NAC vertical
Rancho Seco, TN horizontal
NEI Used Fuel Management Conference, May 6-8, 2014
8
Two categories of casks for dry storage •Bare fuel (also called direct load)
– 11% in 2012
– All metal containers
– Bolted closed
•Canister, thin-walled inner stainless steel container – 89% in 2012
– Overpack of concrete (or sometimes metal)
– Welded closed
•Licensed for up to 20 yr with 20 yr renewal increments
•10 CFR 71
•Current assumption in environmental impact statement (EIS) is that casks will be reloaded after 100 y
IHLRWMC
4/30/2013
9
NRC has approved 34 designs
•Many more versions because of license revisions and amendments
– 5 storage only designs (316 total casks)
– 29 dual-purpose designs (licensed for storage and transportation which started in late 1980s)
•Cask certification mostly based on modeling
•QA program adequate for certification supplemented by observation from an approved aging management program
IHLRWMC
4/30/2013
10
NRC certifies compliance of transportation casks through 3 tests
Impact Puncture
Fire
V V
9 m drop onto unyielding
surface
1 m drop onto 15 cm
steel bar
800 °C fully engulfing fire for
30 minutes
11
Modeling has progressed such that numerical simulations usually sole basis of certification
End
CG over
Corner Side
12
New railcars necessary for transporting massive casks on large scale basis
•Without new railcars, US has no capability to move massive dual-purpose casks
•Association of American Railroads sets the standard for the specialized railcars
•Developing new compliant railcars is long and detailed process of analysis and testing
•DOE currently developing a request for proposals (RFP) to design, test, and certify new railcars
IHLRWM Conference
4/30/2013
13
Dedicated train for rail transportation
Locomotive • Two 4000 HP • Electronically controlled
pneumatic brakes
Cask Car • Carry casks and cradle from 25 to
160 ton • 17 ft long, 12 ft wide, <15 ft tall
Buffer Car • Spread axle loads for bridges • Provide distance to protect
locomotive and escort car • Carry spare parts
Escort Car • Carry security and technical
personnel • Provide location monitoring, and
security/emergency communications
14
Concern for transportation route as great as concern for siting a consolidated storage facility
If storage / transportation route for SNF was proposed within 50 miles of your residence, how likely is it that you would …
Likelihood of Activities (1 = Not At All Likely—7 = Extremely Likely)
Interim Storage
Transportation Route
Attend informational meetings held by authorities (E75/T)
4.37 4.22
Write or phone your elected representatives (E78S/T) 4.20 4.24
Express your opinion using social media (E77S/T) 3.96 4.02
Serve on a citizens’ advisory committee (E81S/T) 3.92 3.91
Help organize public support (E80S/T) 3.07 3.09
Help organize public opposition (E79S/T) 3.05 3.10
Speak at a public hearing in your area (E76S/T) 2.97 3.08
Means
15
Public comments on National Transportation Plan for SNF ask for full-scale testing to address risk concerns
Sandia truck cask test at 130 km/h in 1978
BAM CASTOR side impact test (BAM public website)
16
Possible full-scale testing
•NRC recommendations – Impact test of a rail cask into an unyielding target at 96 to
144 km/h (60 to 90 mph)
– “Back breaker” impact test of a truck cask onto a rigid semi-cylinder where impact limiters are by-passed and the full impact of the test is on the cask itself
– Fully engulfing fire tests for a duration beyond the 30 minute limit specified in 10 CFR 71.73
•National Academy of Science recommendations – Very long duration fire test with a well-instrumented
package to provide validation-quality data
– Regulatory and credible, extra-regulatory impact testing to support integrated analytical, simulation, and scaled testing efforts
17
Stranded SNF storage at shutdown nuclear reactors big issue
•Costs of storing SNF at a shutdown reactor are large and provide large impetuous for consolidated interim storage facility
•Prior to 2000, focus of cost comparisons were between – (a) at-reactor storage (at operating reactor) then repository disposal and
– (b) consolidated interim storage then repository disposal
•By 2013, at-reactor storage had been implemented but a repository was far in the future
•By 2013, focus of cost comparisons were between – (a) at-reactor storage followed by stranded storage then repository disposal and
– (b) at-reactor storage followed by storage at consolidated interim storage then repository disposal
IHLRWMC
4/30/2013
18
Combined cost of storage at reactor followed by stranded storage was ~$35 billion in 2012
• Annual cost for storage is 10 greater at shut down site versus operating site (i.e., ~$6 million/y versus ~$0.6 million/y)
• Costs increase around 2035 when many reactors shut down
• Cost has increased to ~$50 billion based on higher costs for preparing fuel for storage and annual costs for storage at shutdown reactors
19
Consolidated interim storage is path to integrating US waste management system
Consolidated interim storage facility could •Facilitate more flexible siting criteria by implementing schemes to lower
thermal output by – Buffer storage of hot canisters, or – mixing SNF fuel in disposal canister
•Ease burden of aging inspections at shutdown sites and operating sites •Accommodate shipment of bare fuel in wet storage •Make same national organization responsible for long-term storage and
disposal (versus current scheme of private utilities for storage and federal government for disposal)
Consolidated interim storage facility way for the US waste management system to be more flexible to changing situations (e.g., different repository media, emergency closure of reactor, and temporary closure of repository for upgrades)
IHLRWMC
4/30/2013
20
Blue Ribbon Commission on America’s Nuclear Future Reviewed the Back End of the Cycle
•Emphasized Interim Storage as Part of an Integrated Waste Management System
•Consolidated Storage would… – Allow for the removal of ‘stranded’ spent fuel from shutdown reactor sites
– Enable the federal government to begin meeting waste acceptance obligations
– Provide flexibility to respond to lessons learned from Fukushima and other events
– Support the repository program
– Provide options for increased flexibility and efficiency in storage and future waste handling functions
•The Administration agrees that interim storage should be included as a critical element in the waste management system
•The Administration supports a pilot interim storage facility initially focused on serving shut-down reactor sites.
World Institute for Nuclear Security, June 10-12, 2014 Jeff Williams
21
• Accept dry storage containers from stranded sites • Transport fuel dual purpose canisters (DPC) in approved transportation overpack casks • Transfer the DPC to a new storage overpack cask approved for each DPC • 9 stranded sites use 13 canister designs, 8 storage, and 7 transport overpack designs
– Transition from short-term storage to transportation to long-term storage – Aging Management Plans expected
Pilot Storage Facility Concept
• 5,000 to 10,000 tonne capacity with a receipt rate of 1,500 tonne/y
Facilities will include: • Rail yard and associated maintenance equipment • Cask-handling building for transfer of the DPC from
transportation to storage overpacks • Storage pads with multiple vertical and horizontal
storage overpack designs • Security facilities • Infrastructure and balance of plant facilities
World Institute for Nuclear Security, June 10-12, 2014 Jeff Williams
22
Pilot Alternative Design (Flexible, Adaptable, and Expandable)
Dry Storage Alternatives • Vented concrete at grade in horizontal and vertical vendor specific systems currently in use • Vaults for dry canisters • Universal storage overpacks • Universal underground systems
Required Support Systems/Facilities • Cask-handling facility
– large shielded cell vs. transfer cask may offer time in motion and ALARA advantages • Storage overpack fabrication • Rail and cask maintenance • Security systems, infrastructure, and balance of plant
Potential Co-located Systems • Laboratory for supporting long-term storage and developing repackaging techniques • Fuel remediation capability for damaged or failed fuel • Related manufacturing facilities
Humboldt Bay Underground Storage
World Institute for Nuclear Security, June 10-12, 2014 Jeff Williams
23
Larger ISF Concept
Pool
44,000 MT
Dry 48,200 MT
2024 Projected Inventory
DOE Strategy document provides guidance • ISF starts operations in 2025 • 20,000 tonne or greater • Receipt rate is greater than the U.S. discharge rate (~2000
tonne/y), working basis is 3,000 tonne/y • Repository starts operation in 2048 • Modular approach for functional capability and capacity
increases and provide flexibility Assummed ISF capacity is about 70,000 tonnes
• Based on 3,000 tonnes/y receipt rate and schedule in DOE Strategy (2048 repository)
Continued DPC storage using the storage method selected for the Pilot
Significant bare fuel receipt and storage capability may be needed for efficient acceptance from reactors
Pilot and ISF licensed as ISFSI (10 CFR 72)
23
World Institute for Nuclear Security, June 10-12, 2014 Jeff Williams
24
For Full ISF Design Bare Fuel Storage May be Included
Bare fuel receipt and storage systems • Pools – technically mature, but expensive
– Choice for Central Interim Storage in Sweden (CLAB) • Continue to load dry canisters
– decay heat per package may limit transportation and disposal
– DPC may become LLW if repackaging for disposal is required
• Vaults – approach used in Spain
Dry storage continues using technologies selected for the Pilot Support facility capacity increases
• Examine a range of receipt rates Potential packaging facility to sup
disposal if required
24 World Institute for Nuclear Security, June 10-12, 2014 Jeff Williams
25
Why has Germany abandoned Consolidated Interim Storage?
•Transportation risks have been cited, but how extensive was the public discussion?
•Will the prospect of 80 y long term storage cause Germany to re-examine decision?
Perspectives on Dual-Purpose Canister Direct Disposal Feasibility Evaluation
E.J. (Tito) Bonano, E.L. Hardin and E.A. Kalinina
Sandia National Laboratories Albuquerque, NM
SNL/BAM Collaboration Workshop
October 6-8, 2014
Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-AC04-94AL85000.
Unclassified-Unlimited Release (SAND2014-3482C)
Acknowledgments
Justin Clarity, Rob Howard, Josh Jarrell, Eric Pierce & John Scaglione – Oak Ridge National Laboratory
Joe Carter & Tom Severynse – Savannah River National Laboratory
Mark Nutt – Argonne National Laboratory
Christine Stockman – Sandia National Laboratories
Bob Clark – U.S. Department of Energy, Office of Used Nuclear Fuel Disposition
2 Bonano et al., NEI Used Fuel Management, May 2014 (SAND2014-3482C)
Context This is a technical presentation that does not take into account the contractual limitations under the Standard Contract. Under the provisions of the Standard Contract, DOE does not consider spent fuel in canisters to be an acceptable waste form, absent a mutually agreed to contract modification.
3
4
Dry Storage Projections (TSL-CALVIN)
2035: > 50% of commercial used fuel in the U.S. will be stored in ~7,000 DPCs 1,900 canisters now, >10,000 possible 160 new DPCs (~2,000 MTHM) per year At repository opening (2048) the oldest DPC-fuel will be >50 years out-of-reactor Reactor and pool decommissioning will accelerate transfers to DPCs
Bonano et al., NEI Used Fuel Management, May 2014 (SAND2014-3482C)
20-year reactor-life extensions No new builds
Q: Why evaluate technical feasibility of direct disposal of large dual-purpose canisters?
A: Potential for Less fuel handling Simpler UNF/SNF management (facilities, siting, etc.) Lower cost
Re-packaging cost (operations, new canister hardware) 10,000 waste packages for U.S. SNF vs. up to 9X that many for
smaller packages Lower worker dose Less waste (e.g., not disposing of existing DPC hardware)
Technical Evaluation of DPC Direct Disposal Feasibility
Bonano et al., NEI Used Fuel Management, May 2014 (SAND2014-3482C)
Key Technical Assumptions for DPC Direct Diposal Feasibility Evaluation
Complete disposal operations (i.e., panel closure) at/before fuel age of 150 years from reactor discharge
DPC-based waste package size: 2 m dia. × 5 m long, and 80 MT
Waste package + shielded transporter: ≥ 175 MT
Fuel and canister condition will be suitable for transport and disposal for 100 years from reactor discharge
DPCs will be placed in disposal overpacks
Regulatory context for disposal similar to 40CFR197 and 10CFR63
Low probability and low consequence arguments may both be used to evaluate criticality
6 Bonano et al., NEI Used Fuel Management, May 2014 (SAND2014-3482C)
7
Path to Direct Disposal of Existing Storage-Only and Dual-Purpose Canisters
Bonano et al., NEI Used Fuel Management, May 2014 (SAND2014-3482C)
8
Engineering challenges are technically feasible Shaft or ramp transport In-drift emplacement Repository ventilation
(except salt) Backfill prior to closure
SALT
DPC Direct Disposal Concepts
Bonano et al., NEI Used Fuel Management, May 2014 (SAND2014-3482C)
Source: Hardin et al. 2013. FCRD-UFD-2013-000171 Rev. 0.
Time to Repository (Panel) Closure for Representative Disposal Concepts
Based on: Hardin et al. 2013. Collaborative Report on Disposal Concepts. FCRD-UFD-2013-000170 Rev. 0.
32-PWR size packages
Clay/shale concept and any backfilled
concept require much longer aging
Hard rock concept (unbackfilled,
unsaturated, with small and large
spacings)
9
Salt concept (backfilled; 30 m WP, 30 m drift spacing)
Sedimentary (unbackfilled; 30 m WP, 100 m drift spacing)
Hard rock open (unbackfilled; 10 m WP, 70 m drift spacing)
Hard rock open (unbackfilled; 20 m WP, 70 m drift spacing)
Salt concept
Bonano et al., NEI Used Fuel Management, May 2014 (SAND2014-3482C)
Analysis of Postclosure Criticality - Summary
10
Loss of Absorber & Structural
Degradation
Moderator Displacement
& Chloride Brine
Generic burnup credit 32-PWR canister (cask) PWR fuel (4% enriched, 40 GW-d/MT burnup) Original Figure: Wagner J.C. & C.V. Parks 2001. NUREG/CR-6781, Fig. 3.
Bonano et al., NEI Used Fuel Management, May 2014 (SAND2014-3482C)
Stylized Postclosure Criticality Event Tree
11
Original chart from Scaglione et al. 2014. Criticality Analysis Process for Direct Disposal of Dual Purpose Canisters. ORNL/LTR-2014/80. Oak Ridge National Laboratory.
Ground Water
Fresh
Flooding
Rapid Absorber Corrosion (e.g., Boral)
SS Basket Rate << Absorber
Zircalloy Rate << Absorber
Corrosion Rates:
Chloride Brine
Dry
Containment Integrity
Slow
Rapid
Rapid Modify with siting and overpack
functionality
Bonano et al., NEI Used Fuel Management, May 2014 (SAND2014-3482C)
Possible DPC Direct Disposal, Re-Packaging and STAD Canister Strategies
12
STAD Canister ≡ Storage, Transport and Disposal, Multi-Purpose Canister
Existing Canister Designs New Design
Storage-Only Canisters:
Re-Package→ Disposal
DPCs: Re-Package→
Disposal
DPCs: Direct
Disposal
Operational Switch to STAD
Canister at Power Plants
1. No near-term changes→ Re-package (current path) √ √
2. No near-term changes→ Maximize direct disposal (evaluate)
? √ 3. Multiple modes of disposal→
Minimize re-packaging (evaluate)
? √ √
4. Re-package→STAD canister full implementation √ √ √
Bonano et al., NEI Used Fuel Management, May 2014 (SAND2014-3482C)
Fuel Age at Emplacement in a Repository Compared to Re-Packaging in Small STADS
13
Plots show disposition of ~140,000 MTHM U.S. SNF – For 10 kW limit, emplacement could be mostly complete by 2130 – Smaller canisters accelerate disposal but SNF age at disposal is similar
Calculated using TSL-CALVIN (DRAFT)
Bonano et al., NEI Used Fuel Management, May 2014 (SAND2014-3482C)
Timing of DPC Direct Disposal Compared to Re-Packaging in Small STADS Sensitivity Case: Accelerate Repository Opening to 2036
14
Limiting Fuel Age at Disposal is Sensitive To: – Smaller canisters for earlier cooling to emplacement limits – Earlier repository opening date to take advantage of earlier cooling
Calculated using TSL-CALVIN (DRAFT)
Bonano et al., NEI Used Fuel Management, May 2014 (SAND2014-3482C)
15 Bonano et al., NEI Used Fuel Management, May 2014 (SAND2014-3482C)
All options for DPC direct disposal are not the same: Thermal Management
– Favors salt, hard-rock open concepts Size and Operations
– Repository area ranges from 500 to 3,000 m2/package, with zero to 100 years of repository ventilation
– Favors salt and hard-rock open concepts Postclosure Criticality
– Favors salt and very dry unsaturated settings Human Intrusion
– Generally favors crystalline or hard rock
Therefore, waste packaging decisions (such as continued DPC use with the intention of direct disposal) could impact disposal system
design and technical criteria for site evaluation.
16 Bonano et al., NEI Used Fuel Management, May 2014 (SAND2014-3482C)
What are some important implementation risks associated with DPC direct disposal?
Licensing Complexity: Safety analysis could require separate, conclusory calculations for >20 canister types (e.g., criticality calcs.) or even separate calcs. for each as-loaded canister.
Documentation: Utilities would need to produce data on fuel condition and loading, especially for as-loaded postclosure criticality analysis of degraded canisters.
Verification: Canister QA/QC (as performed by utilities and vendors) to include mis-load probabilities, could be important.
Criticality Consequence Analysis: For disposal environments with fresh groundwater, criticality consequence analysis could be needed.
Siting: Some geologic settings could involve more complex analysis to understand DPC-based waste package performance
Preliminary Technical Evaluation of DPC Direct Disposal Alternatives: Summary and Conclusions Disposal Alternatives
– Thermal, criticality, and engineering challenges were identified – Disposal concepts for salt, clay/shale and hard rock were developed
Thermal Results – Repository (panel) closure possible for fuel age < 150 yr – R&D needs have been identified for concepts where clay-rich
materials could see peak temperature > 100°C Preliminary Logistics Results
– At 10 kW thermal limit, emplacement could be complete at 2130 with average throughput of 1,700 MTHM/yr
– To significantly decrease fuel age at emplacement, early repository opening and STAD implementation (smaller canisters) are needed
17 Bonano et al., NEI Used Fuel Management, May 2014 (SAND2014-3482C)
Preliminary Technical Evaluation of DPC Direct Disposal Alternatives: Summary and Conclusions, cont.
Preliminary results indicate DPC direct disposal could be technically feasible, at least for certain concepts. Cost savings could be realized
compared to re-packaging, and further analysis is underway.
Criticality Scoping Results – “Extra” reactivity margin is available using burnup credit analysis
with as-loaded assembly information – Preliminary results show some, but not all, DPCs could be sub-
critical for the degraded cases defined – Saline water (35Cl > seawater) could provide significant neutron
absorption
18 Bonano et al., NEI Used Fuel Management, May 2014 (SAND2014-3482C)
Postclosure Criticality – Prevalence of high-chloride groundwaters in different geologic settings – In-package canister/basket degradation, chemistry and configuration model – Overpack reliability
Waste Isolation/Performance Assessment – System models that discern DPC vs. purpose-built canister performance – Supporting process models for thermally driven coupled processes
Concept Development & Thermal Management – Cavern-retrievable or vault-type concept development – High-temperature backfill (→ 200°C) – Sinking of heavy packages in plastic media such as salt and claystone
Engineering Feasibility, Operational Safety & Cost
Fillers
DPC Direct Disposal Feasibility Evaluation Technical R&D Priorities:
19 Bonano et al., NEI Used Fuel Management, May 2014 (SAND2014-3482C)
Energy 2014: 1
Public Acceptance and Preferences for Used Nuclear Fuel Management
in the U.S. Hank C. Jenkins-Smith
Kuhika Gupta Center for Energy, Security & Society
University of Oklahoma
Energy 2014: 2
Research Goals and Methods ♦ CES&S: Partnership between University of Oklahoma and
Sandia National Labs ♦ Goal: track and analyze the evolution of public perceptions
about UNF management in the U.S. ♦ Methods: complementary streams of data, such as:
• Public opinion surveys » Annual surveys since 2006, totaling > 19,000 participants » Latest survey fielded on 27-28 June 2014, n=1,610
• Social media and big data platforms » Analyzing the co-evolving public and elite narratives using data
collected via Twitter, Google News, and Google Trends • Qualitative focus groups
» Studying group deliberation to assess the kinds of information that stakeholders would like to see when evaluating a prospective UNF management policy
Energy 2014: 3
Research Goals and Methods ♦ CES&S: Partnership between University of Oklahoma and
Sandia National Labs ♦ Goal: track and analyze the evolution of public perceptions
about UNF management in the U.S. ♦ Methods: complementary streams of data, such as:
• Public opinion surveys » Annual surveys since 2006, totaling > 19,000 participants » Latest survey fielded on 27-28 June 2014, n=1,610
• Social media and big data platforms » Analyzing the co-evolving public and elite narratives using data
collected via Twitter, Google News, and Google Trends • Qualitative focus groups
» Studying group deliberation to assess the kinds of information that stakeholders would like to see when evaluating a prospective UNF management policy
Energy 2014: 4
Preferred Energy Sources
2006–2014
Renewables: 0.0% Fossil: + 27.6% Nuclear: – 31.8%
29 25 27 25
33 36 35 37
22 23 22 23 20
17 16 15
49 52 51 52
47 47 49 48
0
10
20
30
40
50
60
2006 2007 2008 2009 2010 2011 2012 2013 2014
%
Fukushima
(e18A, 19A, 20A)
What percent of our energy should come from nuclear energy, which currently provides about 8% of total U.S. energy?
Energy 2014: 5
As nuclear fuel is used to generate electricity, it becomes contaminated with radioactive byproducts. When it can no longer efficiently produce electricity, it is called “used” or “spent” nuclear fuel. To the best of your knowledge, what currently is being done with most of the used nuclear fuel produced in the U.S.? (response options randomized)
22 24 23
25
32 41
39 39
35
36
30
34 32
29 25
22
23 24
13 14 16 17
15 12
15 15 17
29 32
27 26
24 23
24
23 24
0
5
10
15
20
25
30
35
40
45
2006 2007 2008 2009 2010 2011 2012 2013 2014
Cooling pools or special storage containers at nuclear power plants
Regional storage sites
Chemically reprocessed and reused
Nevada repository
%
(e33)
Knowledge about UNF Policy
Energy 2014: 6
Current On-Site Storage
♦ Move radioactive materials only once to permanent repository
♦ Packing & transporting materials to ISF is risky
♦ Less expensive in short-term; buys time for permanent solution
♦ No harm yet; risks of terrorism and flooding can be reduced
♦ Improving protections from terrorism and flooding expensive
♦ Near large populations; UNF has leaked into pools
♦ Quantities of UNF increasing with no permanent solution
♦ UNF at “stranded” sites expensive to secure and protect
Arguments FOR Arguments AGAINST
Strongly Oppose Strongly Support
Mean = 3.57 (e35)
% 14 11
18
31
16
6 4
0
5
10
15
20
25
30
35
1 2 3 4 5 6 7
Energy 2014: 7
Interim Storage
♦ Construct sooner than repository; store UNF up to 100 yrs.
♦ Better protection from terrorists; allows packaging for repository
♦ Reduce UNF storage near pop. centers; reduce risks of flooding
♦ Eliminate stranded fuel; savings help offset costs of ISF
♦ Could delay decision on permanent disposition
♦ Risks of transportation > risks of on-site storage
♦ Cheaper & politically more acceptable than new facilities
♦ No public harm yet; risks of terrorism, flooding can be addressed
Arguments FOR Arguments AGAINST
10 8
15
27 23
11 7
0
5
10
15
20
25
30
35
1 2 3 4 5 6 7 Strongly Oppose Strongly Support
Mean = 4.04 (e36)
%
Energy 2014: 8
Proximity to ISF Now assume that this interim storage facility is to be located [50, 100, 150, 200, 250, or 300] miles from your primary residence. (distances randomized)
3.85 (p = .1190)
3.98 (p = .6118)
3.73 (p = .0619)
3.50 (p < .0001)
3.64 (p = .0006)
3.34 (p <.0001)
4.04
1 2 3 4 5 6 7
300
250
200
150
100
50
Not Stated
Strongly Oppose
Strongly Support
(e37)
(e36)
Distance in Miles
Energy 2014: 9
WIPP Incident On the evening of February 14, 2014, trace amounts of airborne radioactive materials were discovered above ground near the facility. It was determined that 21 workers were exposed to trace levels of radiation. No deaths or serious injuries have been reported, and no one is known to have been exposed to harmful levels of radiation. Pictures from the underground facility show the lid of a drum of waste burst open in a room that is partially filled with containers of radioactive waste. An open drum could release radioactive material into the air flowing through the repository. The cause of the burst lid in an unsealed room is under investigation.
Implications of WIPP Incident for Support of ISF
0
10
20
30
40
-10 -9 -8 -7 -6 -5 -4 -3 -2 -1 0 1 2 3 4 5 6 7 8 9 10
Strongly Reduced
Strongly Increased
No Effect
%
Mean = –1.87 (e41)
19% 50%
31%
Energy 2014: 10
Valuing UNF Storage Options Government officials are deciding how to proceed on storing used nuclear fuel in the U.S. Their decision on how these materials should be stored could cost you money. For example:
• Continuing to store used nuclear fuel at nuclear power plants would require heightened security measures and expanding current practices, which is expensive and could mean higher taxes.
• Construction of interim storage facilities and transportation of used nuclear fuel to the facilities is expensive and could mean higher taxes.
Energy 2014: 11
ISF Siting Process: Who Should Have Vito Power?
Select all of the following that you think should be allowed to block or veto construction of a proposed interim storage facility for used nuclear fuel.
A majority of citizens, including those in Native American communities, residing within 50 miles of the proposed facilities 66
A majority of voters in the host state, including affected Native American communities 64 The host state’s environmental protection agency or its equivalent 55 The Governor of the host state 52 The US Environmental Protection Agency 50 The US Department of Energy 44 The US Nuclear Regulatory Commission 43 Either of the two US senators representing the host state 39 The US congressperson representing the host district 39 The leaders of the host state’s legislature 39 Tribal authorities of affected Native American communities 38 Nongovernmental environmental interest groups in the host state 26
%
(e65)
Energy 2014: 12
ISF Siting Process: Likely Modes of Participation
Assuming construction of an ISF is proposed within 50 miles of your residence, how likely is it that you would . . .
Attend informational meetings on the proposed ISF held by authorities (e76)
33 18 50
Contact your elected representatives expressing your opinion regarding the proposed ISF (e79)
38 19 43
Express your opinion on the proposed ISF using social media (e78)
40 16 44
Speak at a public hearing about the ISF (e77) 58 17 25
Help organize public opposition to the proposed ISF 56 20 24
Unlikely
(1–3) Unsure
(4)
Likely
(5–7) (1 = Not At All Likely—7 = Extremely Likely)
Energy 2014: 13
Willingness to Engage: ISF Citizens’ Advisory Committee
If invited, how likely is it that you would participate as a member of a citizens’ committee asked to provide advice and oversight to authorities developing the proposed ISF if it required about [5, 10, 20] hours of your time monthly for a year? (times randomized)
45
18
37 41
18
41 45
20
35
0
10
20
30
40
50
Unlikely (1–3)
Unsure (4)
Likely (5–7)
%
(e82)
5 Hours 3.81
10 Hours 3.83
20 Hours 3.70
Means (1 = Not At All Likely—7 = Extremely Likely)
Energy 2014: 14
Conclusions ♦ Preferences for nuclear in future energy mix have been
declining since Fukushima • But current percentage (8%) is lower than preferred (15%)
♦ Mixed understanding of current UNF management policy ♦ Support for interim storage is higher than support for
current on-site storage • Support for ISF decreases with proximity • WIPP incident has potential to decrease support for ISF
♦ Non-market value of an ISF is higher than non-market value of continued on-site storage • Inclusion of a research lab and repackaging facility increases non-
market value of an ISF ♦ Local residents most likely to have initial NIMBY response
• Substantial fractions of population willing to engage • Absent state level opposition, engagement can reverse NIMBY
Photos placed in horizontal position with even amount of white space
between photos and header
Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-AC04-94AL85000.
Truck Transport Results / Progress on Rail Test SNL-BAM Workshop
7 October 2014 Paul McConnell
SAND2014-18948 PE
What we think we know The strains measured in the test program were in the micro-strain
levels – well below the elastic limit for either unirradiated or irradiated Zircaloy-4.
Based upon the test results, which simulated normal vibration and shock conditions of truck transport, strain- or stress-based failure of fuel rods during normal transport seems unlikely.
Strains on irradiated rods may be less than strains measured on unirradiated tubes.
Normal conditions of truck transport are more severe than rail.
2
ISFSI Locations
3 http://www.enviroreporter.com/wp-content/uploads/2013/10/NRC-map-of-Independent-Spent-Fuel-Storage-Installations.jpg
Lots of Assemblies to be Stored & Transported
4
Transportation
5
ISFSI
Hoosac Tunnel See: http://en.wikipedia.org/wiki/Hoosac_Tunnel
NAC-MPCs (MPC-36 canisters) NAC-STC rail cask
Courtesy Yankee Rowe
Heavy-haul truck required to get to railhead Courtesy Yankee Rowe
Yankee Rowe
Transportation
6
Railhead, Portland, Conn. near Connecticut Yankee
Barge transport, Connecticut River (Connecticut Yankee pressure vessel)
Courtesy Connecticut Yankee
Connecticut Yankee barge slip site
Parking lot for heavy-haul truck access to railhead!
Central Storage Facility
7
http://www.world-nuclear-news.org/WR-Rethink_on_Utah_used_fuel_storage_project-0408104.html
Private Fuel Storage NRC-licensed design Goshute Reservation, Utah
Repository
8
http://sanjindumisic.com/onkalo-spent-nuclear-fuel-repository-future-of-monuments Onkalo Facility, Finland
Not a Repository
9
Ken Sylvia
Us
There Will Be Lots of High Burnup Assemblies
10
Motivation for assembly testing
Federal Regulations require an assessment of “Vibration - Vibration normally incident to transport”… …imposed on transport packages and contents during “normal conditions of transport”. (10CFR71.71)
The NRC has approved normal transport of low burnup spent fuel.
However, the technical community needs to establish a technical
basis to demonstrate that high burnup fuel rods can withstand all normal conditions of transport.
Vibrations and shocks have been measured on truck trailers and railcars but not directly on fuel assemblies, baskets, or fuel rods.
11
In other words, could Zircaloy cladding fracture during normal conditions of transport?
12
http://sanonofresafety.org/nuclear-waste/
Application of Fuel Assembly Test Results (1)
The margin of safety between the applied loads on fuel rods during transport and the material properties of Zircaloy rods has not been quantified. The SNL assembly tests provide data – the applied stresses on the rods - related to the issue of the margin of safety:
applied rod stressnormal transport Material property test programs at other national laboratories shall measure properties of high burnup cladding:
yield strengthcladding
Application of Fuel Assembly Test Results (2)
• The data from the assembly tests will be used to validate finite element models of fuel assemblies.
• The validated models can be used to predict the loads on fuel rods for other basket configurations and transport environments, particularly rail.
FUEL ASSEMBLY SHAKER TEST SIMULATION, Klymyshyn, et al., PNNL, FCRD-UFD-2013-000168, May 2013
SNL Experimental 17x17 PWR Assembly
Isometric View of Fuel Rods (Top Nozzle and Basket not shown)
Only Zircaloy rods were instrumented with strain gauges and accelerometers
Basket/Assembly Test Unit • The test unit included an assembly and a basket. • The basket is based upon the geometry of the NAC-LWT truck cask
PWR basket. • The assembly was placed in a basket which was placed on 1) a
shaker and subsequently 2) a truck trailer. • The assembly had the same freedom of motion within the basket as
it would have in an actual cask.
• 6061 Aluminum Basket • Sides 1.5 inches thick • Top/bottom 1 inch thick • Length 161.5 inches • Weight 837 pounds
Lead Rod within Copper Tube to Simulate Mass of UO2 (Zircaloy-4 tubes also contained Lead)
Copper tube (or Zircaloy rod)
Lead
Uniaxial Accelerometer and Strain Gauge on Test Assembly
18
Spacer Grid Copper Rod
Zircaloy Rod Accelerometer
Strain Gauge
Left: Accelerometers and Strain Gauge on Top-Center Zircaloy Tube and Spacer Grid Right: Assembly within Open Basket. Note the two Zircaloy-4 rods with instrumentation attached
Shaker Shock Test Video Top-end view of assembly in basket
20
Maximum Micro-strains on Zircaloy Fuel Rods during Shaker Shock Test – Strains are very low
21
Maximum Strains on Zircaloy Fuel Rods, Shock Test #1
Rod Location Assembly Span Position on Span Maximum Strain (µin./in.)
Top-middle rod Bottom-end Adjacent to spacer grid 90 Top-middle rod Bottom-end Mid-span 131 Top-middle rod Bottom-end Adjacent to spacer grid 171 Top-middle rod Mid-assembly Adjacent to spacer grid 104 Top-middle rod Mid-assembly Mid-span 97 Top-middle rod Top-end Adjacent to spacer grid 127 Top-middle rod Top-end Mid-span 199 Top-middle rod Top-end Adjacent to spacer grid 70
Top-side rod Bottom-end Adjacent to spacer grid 54 Top-side rod Bottom-end Mid-span 107 Top-side rod Top-end Mid-span 117 Top-side rod Top-end Adjacent to spacer grid 113
Bottom-side rod Bottom-end Mid-span 62 Bottom-side rod Bottom-end Adjacent to spacer grid 121 Bottom-side rod Mid-assembly Adjacent to spacer grid 110 Bottom-side rod Mid-assembly Mid-span 115
Average of All Strain Gages Average Top-middle Rod
Average Top-side Rod Average Bottom-side Rod
Average Bottom-end Span Average Mid-assembly Span
Average Top-end Span Average Mid span
Average Adjacent to Spacer Grid
112 124 98
102 105 107 125 118
107
maximum
average maximum
Test Unit on Concrete Blocks on Trailer
22
Basket/assembly
Concrete simulates mass of a truck cask
Truck Test Route 65 km in Albuquerque area
23
Range of Road Conditions
24
Route included railroad crossings…
25
…and rough dirt roads
26
Strains measured on instrumented rod
27
dip on Area III Access Road Poleline Road Gibson Blvd.
Strains correlated with road conditions
Strains correlated to road surfaces
28
Pennsylvania St. bridge
speeding to Building 6922
8-inch rut
Rod Strains and FFT maximum strains occurred at low Hz
29
Side Basket Showing Cutout for Filming Assembly during Truck Test
30
Langweilig Video of Assembly during the Truck Test
31
Maximum Strains Measured during Truck Test similar to shaker results
32
Strain Gauge Location on Assembly Maximum Micro-strain Absolute Value (µin./in.) Road Segment
S1 - 0° Adjacent to first spacer grid, Span 10
55
1
S1 - 90° 53 S1 - 225° 74
S2 - 0°
Mid-span, Span 10 94
S2 - 90° 99 S2 - 225° 86
S3 - 0°
Adjacent to first spacer grid, Span 5 143
S3 - 90° 84 S3 - 225° 108
S4 - 0°
Mid-span, Span 5 69
S4 - 90° 101 S4 - 225° 93
Average 0° 90
1 Average 90° 83 Average 225° 90
All maximum strains during road Segment #1 at 872.4 – 902.3 seconds into the trip. This corresponds to travel on Poleline Road (dirt).
Measured Strains are Very Low Relative to the Elastic Limit of Zircaloy-4
33
Zircaloy-4 data per Geelhood, PNNL Analysis datum per Klymyshyn, PNNL
MAXIMUM STRAIN TRUCK TEST ≈143 µin./in.
Irradiated rods would experience lower strains during truck test than unirradiated tube Bending stiffness (=EI) of high burnup irradiated Zircaloy-4 with pellet-clad interaction
(per ORNL): EIZirc4-irr = 52 N-m2 (I based upon rod geometry)
EZirc4-irr = 83 – 101 GPa
Bending stiffness of unirradiated Zircaloy-4 tube:
EIZirc4-unirr = 15.9 N-m2 (I based upon tube geometry) E Zirc4-unirr = 99 GPa
Bending stiffness Zircaloy-4 (irradiated rod/unirradiated tube) = 52/15.9 = 3.27
This implies that for a given applied moment, strains on an irradiated rod would be approximately 0.3 (1/3.27) of those on an unirradiated Zircaloy-4 tube.
The maximum strain measured on the Zircaloy-4 tube in the truck test was 147µm/m, so for the same applied loads, the strain on an irradiated rod would be:
147(15.9/52) = 45 µm/m 34
Rail Test Options TN-32 cask transport from Pennsylvania to North Carolina
35
~ 650 km
Rail Test Options
NLI-10/24 cask tests at Tri-City Railroad near PNNL
36
Augusta, Georgia
TCRY Railyard Richland, Washington
37
• Controlled test environment • Variety of track conditions • Repeatability
38
Rail loadings less severe than truck loads
Fracture Mechanics & Fatigue Assessments Based Upon Experimentally-Measured Strains
Crack depth/Zircaloy wall thickness
Applied stress intensity at crack tip, (MPa-√m)
Lower bound Zircaloy-4 fracture toughness, (MPa-√m)
0.10 0.3 20 - 30 0.25 0.4
0.50 0.6
39 Rail cycles
Bernhard Droste 1
SNF/HLW Dual and Multi Purpose Casks Issues
Bernhard DrosteBAM Federal Institute for Materials Research and Testing
Berlin, Germany bernhard.droste@bam.de
BAM/Sandia WorkshopAlbuquerque, NM, USA, October 6-8, 2014
Bernhard Droste BAM/Sandia Workshop 2
Presentation Outline
- Design, Transport, Storage of DPCs for SNF and HLW in Germany
- Measurement and Demonstration Programs
- Integrated DPC Safety Case Approach, IAEA
- Aging Considerations
- Inspections before Transport after Storage
- MPC
Bernhard Droste BAM/Sandia Workshop 3
ca. 38 m
ca. 92 m
SNF and HLW Storage in Dual Purpose Casks
SNF/HLW Interim Storage Facilities
using Dual Purpose casks
(as constructed, built and operated in Germany)
Bernhard Droste 4
DPCs
BAM/Sandia Workshop
Photos: GNS
Dual Purpose Cask for High Level Waste (HLW)CASTOR HAW28MStorage Version
Dual Purpose Cask for Spent Nuclear Fuel (SNF)CASTOR V/19 Transport Version
Bernhard Droste BAM/Sandia Workshop 5
DPC (HLW) Transport Campaigns from France to Germany
Transport of 11 TN85 Casks by Road from La Hague to Valognes, by Rail to
Dannenberg and by Road to Interim Storage Facility Gorleben (2008)
Fotos: NCS
Bernhard Droste BAM/Sandia Workshop 6
SNF and HLW DPC Storage Facility TBL Gorleben
Foto: GNS
Current Inventory:- 108 HLW Casks (1 TS 28V, 74 CASTOR HAW 20/28CG,
12 TN 85, 21 CASTOR HAW28M)- 5 SNF Casks (1 CASTOR Ic, 1 CASTOR IIa, 3 CASTOR V/19)
Bernhard Droste BAM/Sandia Workshop
• CASTOR Ib with 4 PWR SNF Assemblies, NPP Stade-WAK Karlsruhe
• CASTOR Ia with 4 PWR SNF Assemblies, NPP Biblis-KFA Juelich
• TN 1300 with 12 PWR SNF Assemblies, NPP Biblis
• CASTOR Ic with 16 BWR SNF Assemblies, NPP Würgassen
• CASTOR AVR with 2 Stainless Steel Canisters, each filled with 950 spherical „Graphite Ball“ AVR Fuel Elements, KFA Jülich
• TN AVR-2 with the same Contents as before, KFA Jülich
Results:Verification of
- Cask handling operations
- Containment function
- Leakage rates and their measurement methods
- Evacuation, drying and gas filling operations
- Shielding efficiency
- Heat removal
- Fuel rod temperatures
- Fuel rod integrity ; cavity gas sampling
7
German Dry Spent Fuel Storage Demonstration & Measurement Programs with different SNF Dual Purpose Cask Designs:
Dry SNF Storage Demonstration&Measurement Programs 1982-1985
Bernhard Droste BAM/Sandia Workshop
CASTOR Ia with 4 PWR SNF Assemblies09/1983 – 09/1985
Loading at NPP Biblis
Transport and Transfer into a Hall at KFA Juelich
For transportation with a primary lid penetrated byinstrumentation oroficesthe secondary lid needs to be assessed and approved as transport package containment boundary
…that is the same requirement as for storage casks to have a back-up solution in case of a hypo-thetical loss of primary lid`s leaktightness
Dry SNF Storage Demonstration&Measurement Programs 1982-1985
8
Bernhard Droste BAM/Sandia Workshop 9
CASTOR Ia in
Storage Test Position
Thermocouples Penetration
through Secondary Lid,soldered leaktight in small Lid
Dry SNF DPC Storage Demonstration&Measurement Programs
Bernhard Droste BAM/Sandia Workshop 10
Differences between DPC Transport Package and DPC Storage Cask
DPC Storage Package:
- No impact limiters (on the cask)
- Secondary lid/seal with monitoring
- Protection lid
- Vertical position, inside hall
- Acceptance criteria: national storage req.
(e.g. on-site transport and handling accidents)
DPC Transport Package:
- Impact limiters at bottom and lid side,
in some designs also circumferentially
- Transport in horizontal position, under canopy
- Acceptance criteria: SSR-6 (e.g. accident test conditions: 9m drop/1m puncture/30 min fire)
….to be considered in their Safety Cases
2 Dual Purpose Cask configurations:Different acceptance criteria lead to different DPC specifications which have ONE „core assembly“ (contents, basket, body, primary lid)
IAEA Document on Preparation of a DPC Safety Case
WG webpagehttp://www-ns.iaea.org/tech-areas/waste-safety/spent-fuel-casks-wg.asp?s=3
Bernhard Droste BAM/Sandia Workshop 11
Bernhard Droste BAM/Sandia Workshop 12
Design and operational Considerations against Ageing
Design considerations to limit ageing effects (e.g. proper material/component
selection) and operational conditions to limit access of damaging agents (e.g.
drying/evacuation, humidity control) are important issues of safety assessment,
package design and management system approval.
From IAEA-TECDOC-DRAFT“Preparation of a safety case for a dual purpose cask containing spent fuel”
For those components inside the cask and inside the lid closure system, which cannot be changed during the use,it is essential to capture all potential degradation influences at the initial assessment!
Bernhard Droste BAM/Sandia Workshop 13
Cs corrosion tests of the lid closures of 9 small heated containers
Cs corrosion test of Aluminum and Silicon specimen
Investigation of the influence of Cesium on Lid Closure Components
Can Cesium, released from defective fuel rods, cause corrosion of metal seals?
BAM investigations could demonstrate that it is not the case!(1989-1992)
Bernhard Droste 14
Experience in Transport Preparation after Storage
CASTOR© THTR/AVR
BAM/Sandia Workshop
Interim Storage of SNF of decommissioned
gas cooled high temperature research reactor
in Jülich, Germany
© FZJ
� Loaded between 1993 - 2009
� Monolithic ductile cast iron cask body
� Double lid closure system
(permanent pressure monitoring)
� Metallic seals
� Upper & lower pair of trunnions
� Bottom & top impact limiters
(steel sheeted, wood filled)
� 20 years in storage
Bernhard Droste 15
CASTOR© THTR/AVR
BAM/Sandia Workshop
© FZJ
Transport preparation of
152 casks is ongoingExample:
Repair & Testing
of Trunnions
Example:
Leak-Tightness Test
at Primary Lid
© FZJ
Tests and Inspections before Transport after Storage
Preparation for transportationto another destination
Bernhard Droste 16
Test and Inspection Plan for the CASTOR © THTR/AVR Casks
(1) Check of documentation of pressure monitoring system
BAM/Sandia Workshop
����(2) Visual check of surfaces
����(3) Block-Position measurement of all lids ����(4) Examination of bolting torque of primary lid bolts ����(5) Leak-tightness tests of lid systems ( 33 primary lids) ����
(6) All seals of 55 reassembled secondary lids renewed
and leak-tight tested
����
(7) Inspections of bolts and threaded holes (one hole repaired) ����(8) Check of trunnions, refurbished and replaced, 55 casks load
tested ����CASTOR© THTR/AVR fulfills current regulatory requirements
55 packages were inspected and tested
Transport ability was retained after more than 20 years of storage !
Bernhard Droste BAM/Sandia Workshop 17
Conclusions
Essentials for ageing managementof dual purpose transport packages:
1. Design that considers ageing resistance ofcomponents and materials(materials ageing assessment, effective inner and outer coatings and
medium penetration barriers, quality in manufacturing/documentation etc.)
2. Operational conditions that prevent degradation propagationand ingress of corrosive agents as much as possible(drying, evacuation, inert gas atmosphere etc.)
3. Periodic package design approval certificate renewal(gap analysis of the safety case, management system adaption etc.)
4. Inspection program for tests before transport(appropriate selection of measures considering storage experiences etc.)
Bernhard Droste
Drop Test Campaign with a SNF Multi Purpose Cask 1:1 Model (1994)
BAM/Sandia Workshop
POLLUX Cask (GNS)Designed for transport,
storage and disposalof spent nuclear fuel
18
Used Fuel Disposition Campaign
Interim Storage Mock-Up Discussion
David Enos and Charles Bryan Sandia National Laboratories UFD Working Group Meeting June 5th, 2014 SAND2014-15020 PE
Used Fuel Disposition
Background
Considerable work has been done on 304SS to demonstrate that it is susceptible to chloride induced stress corrosion cracking
Work of particular relevance to interim storage relies on bend bars to provide the stress state – Is this representative? – What can these tell us and what are their limitations?
Recall – SCC requires three things – Environment (EPRI work, etc.) – Susceptible material – Mockup (sensitization) – Stress – Mockup (weld residual stress)
2
Used Fuel Disposition
Goals for a Mock Container
Want to replicate fielded structures in order to assess the susceptibility stress corrosion cracking initiation and propagation
Welding parameters, joint designs, etc. are all held proprietary by the vendors
NEUP program (R. Ballinger) approached three vendors last year and received quotes from each of them.
We attempted to do the same with varying degrees of success – NAC – still waiting… – Holtec – no response. – Areva-TN - Ranor
3
Used Fuel Disposition
General Info on the Mock-up
Wall material: 304 SS Wall thickness, overall diameter, weld joint geometry: standard
geometry for NUHOMS 24P Welds:
– Specific design not specified by manufacturer. – Welds to be full penetration and inspected per ASME B&PVC Section III,
Division 1, Subsection NB (full radiographic inspection) – Double-V joint design – Weld procedure: Submerged Arc
4
Used Fuel Disposition
5
Mock-Up Design
67.2
5 in
.
48 in. 48 in.
Two longitudinal welds, 180 degrees apart
Circumferential weld
Used Fuel Disposition
Mock-Up Design
6
67.2
5 in
.
48 in. 48 in.
Three longitudinal welds, 180 degrees apart
Two Circumferential welds
48 in.
Used Fuel Disposition
What do we want to do with the mockup?
Comments on the design – anything we should add/remove? – Baseplate? – Simulated repairs? – Stress mitigation? – Others?
What do we want to measure?
– Weld residual stress state – Extent of sensitization
What samples do we want to make?
– Subdividing the mock-up will impact the stress state – need to determine how much – Sample geometry that we need?
7
Photos placed in horizontal position with even amount of white space
between photos and header
Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-AC04-94AL85000. SAND NO. 2014-18503 PE)
Transportation Logistics Elena Kalinina
BRC and Administration Strategy
2
Emphasized Interim Storage as Part of an Integrated Waste Management System. “Consolidated Storage would... Allow for the removal of ‘stranded’ spent fuel from
shutdown reactor sites. Enable the federal government to begin meeting
waste acceptance obligations. Provide flexibility to respond to lessons learned
from Fukushima and other events. Support the repository program. Provide options for increased flexibility and
efficiency in storage and future waste handling functions”.
“The Administration agrees that interim storage should be included as a critical element in the waste management system.
The Administration supports a pilot interim storage facility initially focused on serving shut-down reactor sites.”
Consolidated Interim Storage Facility (ISF) Concept
3
Pilot ISF (2021) 5,000 to 10,000 MTU. 1,500 MTU/yr receipt rate . Dry storage containers from shutdown sites with
“stranded fuel”. Transport containers in transportation overpack. 9 stranded sites use 13 canister designs, 8
storage, and 7 transport overpack designs.
Full Size ISF (2025) 70,000 MTU or greater. 3,000 to 4,500 MTU/yr receipt rate. Dry storage containers and bare fuel from all the remaining reactor sites:
4 new shutdown sites 100 operating reactor sites
Prepare for the Large-Scale Transportation of Spent Nuclear Fuel (SNF) and High Level
Radioactive Waste (HLW)
4
Collaborating with stakeholders through State Regional Groups and tribal representatives.
Design, testing, and acquisition of rail cars and transportation casks.
Initiate development of S-2043 Compliant Railcars.
Removing SNF from the shutdown reactor sites. Removing fuel from all the reactor sites and DOE
sites.
5
Assist in the process of selecting appropriate strategies for transporting the Spent Nuclear Fuel (SNF) from the shutdown sites.
Explore the logistics and costs associated with shipping SNF to a hypothetical storage facility.
Understand what resources and time would be required to unload the shutdown sites.
Consider possible scenarios of transportation of SNF from the shutdown sites to a potential consolidated storage facility.
Identify major factors affecting scenario performance. Rank (compare) the scenarios based on their performance.
Removing SNF from the Shutdown Reactor Sites
NOTE: the locations of the consolidated storage facilities and the starting date of their operation were selected arbitrarily.
6
Scenario Parameters
1. Campaign duration: 1, 2, 3, 4, 5, 6, and 8 years.
A hypothetical consolidated storage facility starts its operations in 2021.
2. Fuel selection approach: Older Fuel First Sequential unloading when possible. Parallel unloading when possible.
Campaign duration
Fuel selection approach
9 Pickup Schedules (not all the possible combinations)
3. Consist size: 1-car, 2-cars, 3-cars, and site-specific (5 cars for Maine Yankee).
4. Location of a hypothetical consolidated storage facility: SE,SW, NE, and NW.
5. Location of a maintenance facility: co-located and not co-located with the consolidated storage facility. 6. Casks: using NAC-MAGNATRAN instead of NAC-STC casks at Haddam
Neck, Yankee Rowe, and La Crosse sites.
31 different scenarios (not all the possible combinations)
Shutdown Reactor Sites
7
LaCrosse: Nac-MPC MPC canisters
Trojan: TranStor Holtec MPC canisters
Humboldt Bay: Holtec HI-STAR MPC canisters
Rancho Seco: TransNuclear Nuhom canisters
Connecticut Yankee: Nac-MPC MPC canisters
Main Yankee: Nac-UMS MPC canisters
Yankee Rowe: Nac-MPC MPC canisters
Big Rock Point: W150 W74 canisters
Zion: NAC MAGNASTOR TSC canisters
8
Hypothetical consolidated storage facilities: SE – Southeastern USA, SW – Southwestern USA, NE – Northeastern USA, NW – Northwestern USA
SW
Shutdown Reactor Site Location
9
Site Fuel Type
Number of Assemblies Storage Canister
Number of
Canisters
Transportation Cask
Big Rock Point BWR 441 W74 7 TS-125
Connecticut Yankee
PWR 1019 MPC-26, 24 40 NAC-STC
Maine Yankee PWR 1434 UMS-24 60 NAC-UMS
Yankee Rowe PWR 533 MPC-36 15 NAC-STC
Rancho Seco PWR 493 24PT 21 MP187
Trojan PWR 780 MPC-24E/EF 33 HI-STAR 100
Humboldt Bay BWR 390 MPC-80 5 HI-STAR 100
La Crosse BWR 333 MPC-LACBWR 5 NAC-STC
Zion 1 and 2 PWR 2226 TSC-37 61 NAC-MAGNATRAN
Total 7649 247
Shutdown Site Inventory
10
TSL User Interface
and CALVIN
OR
NL
Fire
wal
l
Web Services
TOM
Database User’s Machine Web Server
Application Server
Database Server
TOM - Transportation Operations Model: Models transportation operations. Calculates transportation fleet. Calculates transportation costs.
TSL - Transportation Storage Logistics Model: Generates pickup schedule. Calculates all costs, except
transportation costs. Includes database with the
UNF projection, reactor site information, and cask information.
TOM Database: Cask data. Processing times. Costs (casks, transportation,
security, maintenance and other).
Logistical Simulation Tool TSL-CALVIN
11
La Crosse
Calculated Route from LaCrosse to a Hypothetical Storage Facility in SE
The duration of each trip is calculated based on the transportation routes. Assumption: The transportation networks in the future will be the same as they are now.
12
The following activities are simulated: Traveling to the pickup site. Loading the fuel into casks and onto the transportation asset. Traveling to the storage facility. Unloading the cask, unloading the fuel, and loading the empty
cask onto the transportation asset. Traveling to the cask maintenance facility. Performing cask maintenance. Traveling to the fleet maintenance facility. Performing fleet maintenance.
Transportation Cycle in TOM (begins and ends at the fleet maintenance facility)
There can only be one consist loading at the reactor at a time. The unloading capability at the consolidated storage facilities is
unlimited.
Assumptions:
13
Strip Packing Problem: Scheduling trips for a given year consists of fitting trips into a container.
The individual items (trips) are packed into the container to minimize the container height. Minimizing the height becomes an asset-minimization problem.
Scheduling Algorithm in TOM
time to complete the transportation cycle.
consist size TRIP
Assets
one year
14
Transportation Costs in TOM
Barge (if applicable) Crane Heavy haul (if applicable) Mainline rail Security labor Shortline rail Switching fee 180c charges.
Assumption: The calculated mainline rail costs are an approximation of what the actual charges would be. The costs are a function of the weight of the casks, the number of cask cars, and the distance travelled.
Capital Costs Maintenance Costs Operational costs
Purchase Buffer Railcar Purchase cask Purchase Cask Railcar Purchase Escort Railcar
Annual cask maintenance Escort fleet maintenance Standard cask maintenance Transport fleet maintenance
15
Campaign Duration Scenario parameters: parallel schedule, storage in SE USA, co-located maintenance facility.
The high total cost of the short duration campaigns is due to the high capital costs. The 2-car scenarios have higher operational costs (more trips per year), but lower
capital costs (fewer casks).
16
Total Transportation Costs Compared to Dry Storage Costs
Dry utility costs are the costs to maintain dry storage facilities at the remaining shutdown sites.
Dry costs are calculated for the duration of the campaign starting from the first campaign year.
The annual cost of 6 million dollars per site from the CALVIN database was used.
Unloading of the shutdown sites in 3-5 years is optimal with regard to keeping low transportation costs and dry storage costs.
17
Consist Size Scenario parameters: parallel schedule, storage in SE USA, co-located maintenance facility.
The scenarios with the lowest total cost are the ones with the 2-car consists. The number of trips decreases and the trip cost (mostly mainline rail cost)
increases with the consist size.
18
Sequential versus Parallel Approach
The total cost is significantly higher in the sequential approach because more casks are required.
The greater the consist size, the larger the impacts of sequential unloading on the total cost
Scenario parameters: 6-year campaign, storage in SE USA, co-located maintenance facility.
19
Scenario parameters: 2-car consist size, parallel fuel selection approach, and 4-year campaign.
Consolidated Storage and Maintenance Facility Locations
Location ID: 1 –SE, 2 – NE, 3 – SW, 4 -NW
Consolidated Storage Location: The total cost in the case of storage
facility in NW location (farther from the majority of the shutdown sites) is 43% higher than in the case of SE location.
The increase in total cost is due to the increase in operational costs.
Maintenance Facility Location: The total cost in the case of
maintenance facility (NW location) located away from the storage facility (SE location) is 35% higher than in the case when they are co-located (SE location).
The increase in total cost is mainly due to the increase in operational costs.
20
Use of MAGNATRAN Casks
Site-Specific Casks: site-specific NAC-STC casks were used at Haddam Neck, Yankee Rowe, and La Crosse sites. MAGNATRAN: NAC-STC casks at Haddam Neck, Yankee Rowe, and La Crosse were replaced with NAC-MAGNATRAN casks.
Scenario parameters: parallel approach, 2-car consist size, consolidated storage in SE and co-located maintenance facility.
Using the same cask types (NAC-MAGNATRAN) at multiple sites has benefits only for the long duration (greater than 6 years) campaigns.
If the campaign is short, using the same cask type results in higher total costs because some of NAC-MAGNATRAN casks are acquired later in the campaign at the higher price.
21
Scenario Ranking Based on Their Performance
Base Case Scenario: parallel schedule, 4-year campaign, 2-car consist, co-located storage and maintenance facilities in SE.
Capital Costs: The major factor is the campaign duration. The next two important factors are the fuel selection approach and the consist size. Operational Costs: The major factor is the location of the consolidated storage and maintenance facilities. The next important factor is the consist size.
22
Characteristics of the scenarios with highest transportation costs are: Short duration campaign Sequential schedule Consolidated storage located far from the majority of the shutdown sites (NW or SW) Large consist size Maintenance facility not co-located with the storage facility. Characteristics of the scenarios with the lowest transportation costs are: - 4 or 5 year campaign - Parallel schedule - Consolidated storage facility close to the majority of the shutdown sites (NE or SE) - 2-car consist - Maintenance facility co-located with the storage facility. - Site-specific transportation casks (the ones currently licensed for each site). Longer campaigns would be slightly less expensive, but would result in higher dry storage
maintenance costs. The major contributors to the total cost are capital cost and operational cost. Generally, the factors that minimize capital costs (small consist), maximize the
operational costs and vice versa.
Conclusions
NOTE: These result should be used as a general guidance. There are many specific details not considered in this analysis that may affect the selection of the best strategy in unloading the shutdown sites.
Removing SNF from All the Reactor Sites
23
2021 2048 2048
SNF is transported to ISF starting in 2021 and to a repository starting in 2048.
SNF is transported directly to a repository starting in 2048.
ISF Scenarios No ISF Scenarios
Total Transportation Cost
24
Mean Total Cost: $5.3B (No ISF) and $7.2B (ISF)
Transportation Cost Spending Profiles
25
The additional costs in scenarios with ISF are related to transportation from the reactor sites to ISF during 2021 to 2048.
Example of Acquisition
26
Total Casks
Total Vehicles
Total Cost ($B)
Total Miles
Total Trips
233 80 4.3 1.5E7 7228
154 64 5.0 1.2E7 4878
Photos placed in horizontal position with even amount of white space
between photos and header
Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-AC04-94AL85000. SAND NO. 2014-18482 PE
Disposal Research Activities Kevin McMahon
Sandia National Laboratories
Presented to the SNL-BAM Workshop
October 6-8, 2014 Albuquerque, NM
Outline Disposal Research as part of the Used Fuel Disposition (UFD)
Campaign Focus and Technical Challenges for Disposal Research Disposal Options Being Considered
Deep borehole Crystalline Host Rock Argillite Host Rock Salt Host Rock
Work Supporting Disposal Options Dual Purpose Canisters (DPC) Regional Geology Generic Disposal Systems Analysis International Collaborations
Conclusions
2
UFD Campaign Structure
3
R&D Focus for SNF and HLW Disposal
Provide a sound technical basis for multiple viable disposal options in the US
Increase confidence in the robustness of generic disposal concepts
Develop the science and engineering tools needed to support disposal concept implementation
Three mined repository options (crystalline rocks, argillite rocks, and salt) One geologic disposal alternative: deep boreholes in crystalline rocks
4
Technical Challenges and Opportunities for Disposal Research
Building confidence in multiple repository concepts without site-specific data
Developing tools for characterizing complex natural and engineered systems
Identifying constraints on disposal options E.g., different media pose different thermal limits,
constraining repository design and waste package size
Matching engineered barriers to geologic environments E.g., Alloy-22 packages in oxidizing environments, copper
packages in reducing environments
Opportunities for international collaboration France, Germany, Sweden, Switzerland, Korea, Japan,
China, Czech Republic, Canada, Finland, UK …
Minimum decay storage durations to limit peak PWR waste package surface temperature to 100°C (granite, clay) or 200°C (salt). (Hardin et al., 2011, Generic Repository Design Concepts and Thermal Analysis (FY11), FCRD-USED-2011-000143)
TEM of intrinsic Pu(IV) nano-colloids sorbed to goethite at 25°C for 103 days (Wang et al., 2011; Natural System Evaluation and Tool Development—FY11 Progress Report, FCRD-USED-2011-000223)
5
Deep Borehole Disposal Concept
6
Deep Borehole Disposal Considerations
Multiple factors indicate the feasibility and safety of the deep borehole disposal concept
Demonstration site selection guidelines indicate that large areas with favorable geological characteristics exist in the conterminous U.S.
Groundwater characterization should focus on aspects of the system critical to demonstrating safety of the deep borehole disposal system: Groundwater age and history Salinity and geochemistry Potential for vertical fluid movement Permeability in the host rock and disturbed rock zone Borehole seals integrity and durability
7
Crystalline Host Rock Disposal R&D Objectives
Advance our understanding of long-term disposal of used fuel in crystalline rocks and to develop necessary experimental and computational capabilities to evaluate various disposal concepts in such media.
Focuses Better characterization and understanding
of fractured media and fluid flow and transport in such media
Designing effective engineered barrier systems for waste isolation. Especially, the work will take into consideration the implication of the disposal of dual purpose canisters in crystalline rocks.
Shafts
http://www.bbc.com/news/uk-england-cumbria-21253673
OverpackWaste
Canister
Buffer Layer
Crystalline Rock
Invert (reinforced Buffer Material)
1.02.25 m
8
Crystalline Host Rock Disposal R&D – FY14 Accomplishments A R&D plan was developed for used fuel disposal in crystalline rocks. A total of 31
research topics have been identified. A generic reference case for crystalline disposal media has been established. The capability of a discrete fracture network model was demonstrated using fracture
parameters from a testing site. A thermo-hydrologic-mechanical (THM) model has been applied to an engineered barrier
system. Significant progress has made in understanding radionuclide interactions with buffer and
granitic materials. International collaboration has been actively pursued (e.g., DECOVALEX, KAERI, Sweden
URL).
9
9
Argillite Host Rock Disposal Overall Description
Scope An integrated assessment of various aspects of
nuclear waste disposal research in clay-bearing host rock media: Development of a reference case for argillite Geochemical evaluation of interactions relevant to EBS
materials (clay, metal) under repository conditions: – Thermodynamic modeling and hydrothermal experiments – Thermodynamic database assessment: clay minerals and
sorption Coupled Thermal-Hydrological-Mechanical-Chemical
(THMC) – Development and validation of constitutive relationships
for permeability, porosity and effective stress – Discrete fracture network (DFN) approach for fractures in
argillaceous rock – Transport in clay and clay rock Corrosion modeling For used fuel degradation: Application
to Argillite Rock Environments 10
Argillite Host Rock Disposal Examples of Model Development THM coupled models for clay
International Collaborations: THM Modeling of Underground Heater Experiments
Discrete Fracture Network (DFN) approach for fractures in argillite Excavation damaged zone (EDZ) and natural
fracturing Rigid-Body-Spring Network (RBSN) modeling
approach for mechanical damage
Modeling and experimental investigations on barrier material interactions and stability
ANL Mixed Potential Model (MPM) for used fuel matrix degradation Development towards integration with
performance assessment (PA)
11
MPM Model Concept
Salt Host Rock RD&D: Schematic of Features of a Backfilled Repository Room
12
Brine
Vapor
Hot Granular Salt Consolidation, Constitutive Model and Micromechanics Thermal Conductivity as a Function of Porosity and Temperature
Brine Migration Experimental Studies
Material Interactions In Heated Salt Thermodynamic Properties of Brines, Minerals and Corrosion Products In High Temperature Systems
Laboratory Thermomechanical Testing
Radionuclide Solubility Measurements
13
Brine
Vapor
Total System Performance Assessment (TSPA) Model Development Generic Salt Repository Benchmarking TMHC Model Development/Brine Migration
Salt Host Rock RD&D: Schematic of Features of a Backfilled Repository Room
Salt Host Rock Disposal RD&D Field Studies General Objectives • Develop technology and methodology for rock characterization and
testing • Better understand, model and test relevant processes • Better understand various components of engineering barrier system • Provide quantitative data for safety assessment calculations • Test and optimize full-size repository components and operating
procedures (demonstration) • Optimize repository construction techniques • Training and benchmarking • Promote international co-operation • Build confidence in scientific and technical community • Contribute to public trust and confidence
14
Engineering challenges are technically feasible Shaft or ramp transport In-drift emplacement Repository ventilation
(except salt) Backfill prior to closure
SALT
DPC Direct Disposal Concepts
Source: Hardin et al. 2013. FCRD-UFD-2013-000171 Rev. 0.
15
Dry Storage Projections (TSL-CALVIN)
2035: > 50% of commercial used fuel in the U.S. will be stored in ~7,000 DPCs >1,900 canisters now, >10,000 possible with existing reactor fleet 160 new DPCs (~2,000 MTHM) per year Reactor and pool decommissioning will accelerate fuel transfers to DPCs At repository opening (~2048) the oldest DPC-fuel is >50 years out-of-reactor
20-year reactor-life extensions No new builds
16
Regional Geology
17
We are building a GIS spatial database that combines data for alternative host rocks and other natural and cultural features in order to evaluate and communicate potential siting options
Simple geologic characteristics such as depth to formations may be used to evaluate the potential for repository siting in specific regions
Other relevant formation data (e.g., heterogeneity, geochemistry, permeability) will be added to the database in the future
A web-based interactive tool is planned to communicate basic geologic and siting information for different regions of the U.S.
Comparative Depth of (Bedded) Salt Regional Geology Example
18
Generic Disposal System Analysis
Scope Development and implementation of an enhanced performance
assessment (PA) modeling capability, applicable to a range of disposal options (salt, granite, clay, deep borehole)
Technical Challenges Application of high-performance computing (HPC)-enabled PFLOTRAN
code for efficient simulation of 3D integrated multi-physics (thermal-hydro-chemical (THC)) over a range of spatial scales
Representation of spatially-variable THC-driven source term
19
Generic Disposal System Analysis Continued
Accomplishments Development of generic repository reference cases
salt, clay and granite Conceptual model for complex, integrated THC source term
waste degradation, radionuclide solubility and mobility Probabilistic THC simulations and sensitivity analyses
Salt reference case with spatially-varying waste degradation, decay heat, fluid flow, radionuclide mobilization and transport, coupled biosphere
20
International Collaborations in Disposal Research
21
Key Issues Tackled in Current and Planned Portfolio Near-Field Perturbation Engineered Barrier Integrity Radionuclide Transport Demonstration of Integrated System Behavior
International Cooperative Initiatives DECOVALEX - research collaboration and model comparison activity for coupled
processes simulations (currently 10 partners) Mont Terri - research partnership for the characterization and performance
assessment of a clay/shale formation (currently 15 partners) Colloid Formation and Migration - research investigation of colloid
formation/bentonite erosion, colloid migration, and colloid-associated radionuclide transport (currently 9 partners)
FEBEX - in situ full-scale heater test conducted in a crystalline host rock with bentonite backfill (currently 10 partners)
SKB Task Forces - collaboration in the area of conceptual and numerical modeling of performance-relevant processes in natural and engineered systems (currently 12 partners)
International Collaborations in Disposal Research (continued)
22
Bilateral Collaborations KAERI Underground Research Tunnel (KURT) - in situ borehole
characterization and methods for measuring streaming potential (SP) to characterize groundwater flow in a fractured formation
German Federal Ministry of Economics and Technology (BMWi) - model benchmarking and data exchange for salt repositories at WIPP and Gorleben
MoU between ANDRA and DOE - collaborative work in clay/shale disposal at the LSMHM Underground Laboratory near Bure
Countries With Collaboration Partners Include Finland, Sweden, France, Belgium, Peoples Republic of China, Switzerland, Japan, Canada, United Kingdom, Germany, Republic of Korea, Spain, Republic of China (Taiwan)
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