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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

Presenter

Presentation Notes

Recent data from France and Japan shows that at the temperatures of typical dry storage, after about one year, most of the radiation damage in the cladding is thermally annealed, thus restoring much of the ductility of the cladding.

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


High Burn-up Confirmatory Data Project: Rod Testing to Establish the Baseline


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

Presenter

Presentation Notes

High burnup cladding will only become brittle if these conditions are met: 1) the appropriate concentration and distribution of hydrides exists in the as-irradiated cladding, 2) the temperature during drying or storage is high enough to dissolve a sufficient quantity of the original hydrides, and 3) the internal rod pressure is high enough to cause a sufficient hoop stress that enough hydrides reorient radially. Even then, the cladding only becomes brittle once the temperature drops below the DBTT. Above that temperature, the cladding is still ductile.

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

Presenter

Presentation Notes

By providing the realistic external loads on the various Structures, Systems, and Components (not just the cladding), we can determine through the experimental and analytic tasks how much degradation can occur and still have the SSCs meet their safety functions when subjected to these loads and forces. This provides the “end point” to know when we have obtained sufficient data and analyses.

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



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 !


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



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


Holger Völzke

4



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


Holger Völzke

5



12 On-Site Storage Facilities – 3 Designs

Tunnel Design (NPP Neckarwestheim)

Steag Design

WTI Design


Interim Storage Facility at Isar NPP

Holger Völzke



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


Holger Völzke

7




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




§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




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




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




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




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



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


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



14


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







Deformation [mm]

Lo

ad

[N

/mm

]

Reduction

of restoring

seal forceFr

Reduction of ru

Y2

Y1

Al-seal 150 C

Holger Völzke



15


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



16


(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



17


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



18


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



19


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



20


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



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 [email protected]

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


BAM/Sandia Cooperation 1979


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]


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


Signing of the project agreement DOE/BAM

Paris, PATRAM`98

(Bernhard Droste, Kelvin Kelkenberg)

DOE/BAM Agreement 1998, RadWaste Transport


DOE/BAM/SNL Workshop, Albuquerque, 1999


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


BAM/US NRC Cooperation

Visit of US NRC Commissioner Ms K.L.Svinicki

in Berlin and at BAM Test Site, March 24, 2010


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

Presenter

Presentation Notes

Could drop this slide

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.


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


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


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


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


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

Presenter

Presentation Notes

One low-risk path forward would be to implement a standardized multi-purpose canister in the next decade or so, so that DPC direct disposal would be limited to a few thousand canisters.

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


Presenter

Presentation Notes

The costs alluded to here would be borne by industry and the disposal implementor (as is well known to the NEI meeting audience). Disposal costs could also be lower because DPC-based waste packages would be large, thus avoiding hardware and handling costs associated with very many packages (e.g., > 80,000 in the U.S. for 4-PWR size). Cost for disposing of each DPC hull has been estimated to be from $50k to $250k.

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


Presenter

Presentation Notes

Commercial spent fuel will actually be produced in the U.S. over a period of at least 90 years, so we describe the timing of disposal in terms of years since discharge. The size and weight of the largest DPC-based waste packages would be only a little greater than the heaviest packages that were planned for disposal at Yucca Mountain. Disposal overpacks could be corrosion resistant (e.g., copper, titanium, or Alloy 22) or they could be corrosion-allowance (e.g., low-alloy steel) depending on the safety strategies for different geologic settings.

7

Path to Direct Disposal of Existing Storage-Only and Dual-Purpose Canisters


Presenter

Presentation Notes

These are high-level technical questions that would be answered before any particular DPC could be disposed of directly. The dashed outline signifies that these questions could be asked for each DPC (of thousands), to identify a subset that could be suitable for direct disposal. This slide uses the terms thermal limits and criticality requirements, assuming that they are known, but that is not the case today. Requirements will depend on site characteristics and the safety strategy for the disposal system.

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


Source: Hardin et al. 2013. FCRD-UFD-2013-000171 Rev. 0.

Presenter

Presentation Notes

Engineering challenges can be readily met, although some facilities (e.g., waste shaft hoist and/or transporter) would be the largest of their kind. Scalable technologies exist for transporting and waste packages underground. Current methods for lifting, packaging, sealing, shielding, transporting, etc. are within the state-of-the-practice. Development costs would be relatively small, for example, spending $100M to develop waste package transport and handling capabilities, would be a very small fraction of the overall repository implementation cost.

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


Presenter

Presentation Notes

As indicated, this chart is based on temperature limits of 100C for clay/shale media and clay-based backfill, and 200C for salt and hard-rock media. The height of the colored line representing each disposal concept is the waste package thermal power limit at closure. The open-mode concepts are ventilated (hard rock and clay/shale, but not salt), and heat removal efficiency is high, so there is no firm power limit for emplacement. However, there may be power limits for handling and transport underground (e.g., 18 kW for YM).

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.


Presenter

Presentation Notes

This figure shows k-effective vs. time, for a generic 32-PWR canister with typical PWR fuel, flooded with water. Three different burnup-credit cases are shown: a set of actinides only (blue), adding a few fission products (brown, in the middle), and a full set of 28 nuclides (red). As noted on the figure, reactivity increases from in-growth of Am-241 and Pu-240, peaking at about 25,000 years. The point of this figure (without showing hypothetical keff values >1) is that we calculate large increases in keff (roughly +25%) with degradation of the neutron absorber plates and basket. For most DPC designs there is not that much “extra margin” even taking as-loaded fuel characteristics and burnup into account. Flooding with chloride brine tends to decrease reactivity, with the effect from saturated brine (6 molal) approaching –25%.

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


Presenter

Presentation Notes

So what does it mean if criticality cannot be excluded on low probability? This event tree is an example which, if populated with site-specific probabilities, could show the probability of one or more critical waste packages in a repository might be significant depending on site characteristics and overpack integrity (including disruptive events). If the probability of one or more critical packages is >0.1, then criticality would be included in a human intrusion assessment under 63.321(b)(1) and 63.342(b). This is a target for evaluation—a nexus we would like to avoid. Low-consequence screening is allowed, following the postclosure criticality approach in the TSAR approved for Yucca Mountain. The incidence of criticality (i.e., # of packages) would depend on the geologic setting and other factors, and we plan to do scoping analyses.

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 √ √ √


Presenter

Presentation Notes

These are 4 general strategies possible for disposition of canistered SNF: 1. This is the current path (consistent with the Standard Contract). It requires that all SNF in dry storage be re-packaged in canisters purpose-built for disposal. 4. Full implementation of a standardized multi-purpose canister (STAD) with re-packaging of DPCs, would have a similar endpoint but could be much more efficient. 2. and 3. These intermediate strategies involve direct disposal of DPCs, either as the dominant approach or in combination with STADs. These strategies are being evaluated in this study.

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)


Presenter

Presentation Notes

Fuel age at disposal could be important, for example, if long-term cladding degradation or DPC containment integrity are at issue. Some CALVIN assumptions: In the STAD case (on the right) we begin to load STAD canisters for dry storage at all sites, 5 years before repository opening. All DPCs loaded before STAD implementation would be disposed directly (i.e., no re-packaging). Existing reactors are projected with 20-year life extensions (last shutdown 2055). DPC types currently in use are projected thru decommissioning Fuel has gradually increasing burnup until 5% initial enrichment. All fuel is canistered prior to shipment from the power plants, so thermal blending is not included in these simulations. The 10 kW emplacement thermal limit works for the salt concept, and the hard rock unbackfilled, unsaturated concept (YM-type)

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)


Presenter

Presentation Notes

This result (with a relatively low 6 kW emplacement power limit) is one of the more striking improvements in earlier fuel age at disposal, that we have analyzed so far. The calculated improvement results from shifting to smaller STAD canisters (4-PWR size) combined with direct disposal. We assumed no re-packaging. The 6 kW emplacement limit works for sedimentary media as well as salt and hard rock (YM-type concept). Repository ventilation (50 to 100 years) would be needed for sedimentary settings, or if a clay-based backfill will be used, until the power output decays to ~1,700 W (this value comes from the KBS-3 concept).


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.

Presenter

Presentation Notes

Our understanding of the effect of chloride brine (vs. fresh water) on postclosure criticality is corroborated by Preliminary Safety Analysis – Gorleben (2013). This is the “VSG” safety assessment recently completed by a team of German institutions. The 500 m2/package figure is for low burnup fuel in salt or hard rock, while 3,000 m2/package is for moderate-to-high burnup fuel in clay/shale. (SNF waste packages with LEU fuel can never go critical in the disposal environment unless flooded with water, regardless of internal configuration.) Human intrusion borehole frequencies are recommended for sedimentary and crystalline settings, in 40CFR191. This guidance applies to all packages, not just DPCs, but if criticality becomes coupled with human intrusion analysis, then the probability of interception could be more important. For example, what if human intrusion (expect ~6 interceptions per sedimentary repository) significantly increases the incidence of packages going critical? The final statement is true for DPC direct disposal but also for any packaging decision (e.g., standardized canisters) that is made prior to repository siting and characterization. The risk for a future implementor tends to increase as more fuel is canistered.


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

Presenter

Presentation Notes

These “implementation risks” would be incurred by the agency or organization responsible for developing a repository. There are estimated to be 26 canister designs currently in use for dry storage, and this will increase with time. The recent GC-859 request for data on UNF did not include detailed as-loaded assembly burnup data (e.g., axial profiles of burnup), because of the effort that would be involved. Hence, we infer that obtaining detailed data as much as 50 years in the future, could be problematic. In this discussion, “fresh” groundwater could have chloride concentration up to or exceeding that of seawater. Salt has superior thermal dissipation properties, thus as we have said previously, SNF waste packages can be emplaced sooner and hotter, and backfilled immediately without long-term ventilation.

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


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


Presenter

Presentation Notes

The “extra” reactivity margin referred to here is in addition to the value of keff that is accepted for regulatory approval (e.g., keff 0.95 is specified in 60.131). For example, canisters licensed for fresh fuel have “extra” margin when analyzed for irradiated fuel with burnup credit. Chloride brine in a salt repository actually may not allow direct disposal of any DPC. We plan to analyze this in FY15 using an approach that would not necessarily require extensive as-loaded assembly burnup data for DPCs.

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:


Presenter

Presentation Notes

Many sedimentary formations were deposited in seawater and retain that porewater chemistry. However, chloride concentration greater than seawater is needed. This leads us to investigate the occurrence of higher strength brines in sedimentary basins (associated with salt beds), very old granites, and other settings. Next-generation performance assessment tools and supporting process models are under development. The cavern-retrievable or vault idea would store DPCs in galleries underground, with natural and engineered barriers around the canisters such that the system can later be licensed for disposal also. Sinking of packages is an old issue (dates to the 1980’s) that can be addressed by collecting more representative creep data.

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%

Presenter

Presentation Notes

50% below 0; 31% at 0; 19% above 0

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)

Presenter

Presentation Notes

Minorities are more likely than the rest of the population; those with at least some college education and more likely than those with no college education; females are more likely than men to advise; income has a positive and statistically significant relationship with advising

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




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 [email protected]

BAM/Sandia WorkshopAlbuquerque, NM, USA, October 6-8, 2014


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


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


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


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)


• 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


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


CASTOR Ia in

Storage Test Position

Thermocouples Penetration

through Secondary Lid,soldered leaktight in small Lid

Dry SNF DPC Storage Demonstration&Measurement Programs


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



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!


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 !


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


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


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


5

Mock-Up Design

67.2

5 in

.

48 in. 48 in.

Two longitudinal welds, 180 degrees apart

Circumferential weld


Mock-Up Design

6

67.2

5 in

.

48 in. 48 in.

Three longitudinal welds, 180 degrees apart

Two Circumferential welds

48 in.


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



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



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

http://www.bbc.com/news/uk-england-cumbria-21253673

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

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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.

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Presenter

Presentation Notes

Engineering challenges can be readily met, although some facilities (e.g., waste shaft hoist and/or transporter) would be the largest of their kind. Scalable technologies exist for transporting and waste packages underground. Current methods for lifting, packaging, sealing, shielding, transporting, etc. are within the state-of-the-practice. Development costs would be relatively small, for example, spending $100M to develop waste package transport and handling capabilities, would be a very small fraction of the overall repository implementation cost.

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

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Presenter

Presentation Notes

One low-risk path forward would be to implement a standardized multi-purpose canister in the next decade or so, so that DPC direct disposal would be limited to a few thousand canisters.

Regional Geology

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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

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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

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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

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International Collaborations in Disposal Research

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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)

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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)

bam workshop

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