Spent fuel characterization for geological repositories Peter Jansson, Uppsala University Presented at the 2015 symposium of the Swedish Centre for Nuclear Technology (SKC) October 9 With contributions from: Alessandro Borella (SCK•CEN) Denis Janin (E.ON Kernkraft) Arjan Koning (NRG (IAEA)) Malte Pettau (E.ON Technologies) Marcus Seidl (E.ON Kernkraft) Henrik Sjöstrand (UU) Jean-Christophe Sublet (CCFE)
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Spent fuel characterization for geological repositories
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Spent fuel characterization for geological repositories
Peter Jansson, Uppsala University
Presented at the 2015 symposium of theSwedish Centre for Nuclear Technology (SKC)
European council decision regarding Euratom research. ⇒ IGD-TP
European council decision 2006/976/EURATOM:
... the emphasis in Euratom research should be implementation-oriented R&D activities on all remaining key aspects of deep geological disposal of spent fuel and long-lived radioactive waste…
… demonstration of the technologies and safety…related to the management and disposal of waste are emphasised.
European council decision regarding Euratom research. ⇒ IGD-TP
The instrument of European Technology Platforms (ETPs) have been introduced by EC to:
• Provide a framework for stakeholders, led by industry, to define research and development priorities.
• Play a key role in ensuring an adequate focus of research funding on areas with a high degree of industrial relevance.
• Address technological challenges that can potentially contribute to a number of key policy objectives which are essential for Europe's future competitiveness.
European council decision regarding Euratom research. ⇒ IGD-TP
2006 → 2007Waste management organisations (WMO’s) in Europe performed a feasibility study:
“Co-ordination Action on Research, Development and Demonstration Priorities and Strategies for Geological Disposal (CARD)”
⇒ Technology platform for final disposal in deep geological formations launched in November 2009.
European council decision regarding Euratom research. ⇒ IGD-TP
IGD-TP’s vision:
By 2025, the first geological disposal facilities for spent fuel, high-level waste, and other long-lived radioactive waste will be operating safely in Europe.
IGD-TP is committed to:
• build confidence in the safety of geological disposal solutions among European citizens and decision-makers;
• encourage the establishment of waste management programmes that integrate geological disposal as the accepted option for the safe long-term management of long-lived and/or high-level waste;
• facilitate access to expertise and technology and maintain competences in the field of geological disposal for the benefit of Member States.
European council decision regarding Euratom research. ⇒ IGD-TP
IGD-TP’s Founding DocumentsVision
Strategic Research AgendaDeployment Plan
IGD-TP’s Exchange ForumExchange of information, advice,
discussions and proposals.
IGD-TP’s Executive GroupInitiates, decides, funds and directs IGD-TP,
establishes Working Groups.
IGD-TP’s SecretariatSupports IGD-TP’s activities, acts as communication and information centre. Follows up the deployment
of joint activities and updates the Master Deployment Plan. Reports to the EC.
Joint Activity Types for
Deploymentaccording to MDP
Terms of Reference of IGD-
TPWorking Groups
Coordination with EC
Euratom Framework Programme
Adapted from the IGD-TP flyer (2015-10-06): http://www.igdtp.eu/index.php/key-documents/doc_download/276-igd-tp-flyer-2014
European council decision regarding Euratom research. ⇒ IGD-TP
Countries participating in IGD-TP as of October 6AustraliaBelgium (BE)Canada (CA)Czech Republic (CZ)Finland (FI)France (FR)Germany (DE)Greece (GR)Hungary (HU)International OrganisationsItaly (IT)Japan (JP)
- “dedicated to identifying the main RD&D issues that need a coordinated effort over the next years in order to reach the Vision 2025“
- “the SRA identifies the Key Topics of RD&D that have the greatest potential to support repository implementation through enhanced cooperation in Europe“
Research prioritization by IGD-TP.
Key Topics defined by the IGD-TP:
1. Safety case
2. Waste forms and their behaviour
3. Technical feasibility and long-term performance of
Some activities within the IGD-TP are “Cross-Cutting”:
• Dialogue with regulators
• Competence maintenance, Education and Training
• Knowledge Management
• Communication interfaces and other activities
supporting information exchange
Research prioritization by IGD-TP.Some activities within the IGD-TP are “Joint Activities”:
1. Waste forms and their behaviour2. Full scale demonstration of plugging & sealing3. Waste forms and their behaviour on C‐144. Monitoring of the environmental reference state5. Safety of construction and operations6. Confidence increase in the safety assessment codes ‐ materials
interactions7. Monitoring programme8. Safety case benchmarking9. Safety case peer review
10. Long‐term stability of bentonite in crystalline environments11. Sharing of knowledge on HLW container materials behaviour12. Adaptation and optimisation of the repository13. Communicating results from RD&D14. Competence, Maintenance, Education and Training15. Nuclear Knowledge Management16. WMOs Information Exchange Platforms
Research prioritization by IGD-TP.
“All relevant stakeholders in Europe (industry, research centres, academia, technical safety
organisations, non‐ governmental organisations...) who endorse the IGD‐TP
Vision are welcome to join the IGD‐TP and contribute to the Exchange Forum (EF).“
About the SPIRE collaboration for spent fuel characterization.
Examples from IGD-TP’s Master Deployment Plan
High priority:1. Safety case: Refinement of the tools used in safety
assessment.
2. Waste forms: Rapid release fraction and matrix
dissolution of high burn-up fuels.
3. Technical feasibility: Long-term stability of bentonite in
crystalline environments.
6. Monitoring: Technologies and techniques
About the SPIRE collaboration for spent fuel characterization.
Examples from IGD-TP’s Master Deployment Plan
Medium priority:1. Safety case: Refinement of methods to make sensitivity
and uncertainty analyses.
2. Waste forms: High burn-up fuels and criticality.
4. Development strategy of the repository: Methodologies
for adaptation and optimisation during the operational
phase.
About the SPIRE collaboration for spent fuel characterization.
Identified needs for spent fuel characterization• Safety requirements on barriers in storage of spent
nuclear fuel need information on the spent nuclear fuel to be placed there:○ Decay heat○ Reactivity○ Gamma radiation○ Neutron radiation○ Nuclide inventory
• This need is common to both short term and long term storage issues.
About the SPIRE collaboration for spent fuel characterization.
Identified needs for spent fuel characterization• Measurement techniques to reliably determine decay
heat, reactivity, gamma- and neutron radiation from the nuclear fuel need to be established when the storage or repository begin operation.
• Models to be used for calculations or predictions must be validated and approved for use when the storage or repository begin operation.
• Uncertainties of all measured and calculated parameters must be quantified.
“Spent fuel characterization Program for the Implementation of (geological) REpositories” - SPIRE
About the SPIRE collaboration for spent fuel characterization.
About the SPIRE collaboration for spent fuel characterization.
SKB + UU:
• Decay heat measurements of fuel with long cooling time.
• Fuel characterization system’s integration in encapsulation facility (CLINK).
• Radiation impact on heat transfer in the geological storage.
About the SPIRE collaboration for spent fuel characterization.
SKB + E.ON + UU:
• Decay heat measurements of fuel with short cooling time.
• Radiation impact on heat transfer and dose in transport and in interim storage facilities.
About the SPIRE collaboration for spent fuel characterization.
SCK•CEN + UU + SKB:
● Development of:○ A Self-Interrogation Neutron Resonance
Densitometry prototype instrument.○ Medium resolution gamma rays spectroscopy using
Cadmium Zinc Telluride detectors.● Measurements on fresh MOX fuel in known conditions.● Measurements on spent fuel on-site (Clab).
About the SPIRE collaboration for spent fuel characterization.
UU + PSI + CCFE:
• Generalize the Total Monte Carlo scheme to all nuclear data uncertainties
• Calculate the decay heat in and burnup of fuel assemblies and estimate and minimize its uncertainties using the TMC methodology.
• Calculate criticality margins and associated uncertainties for both fuel storage and in deep repositories.
• Benchmark calculation results against both differential (e.g. IGISOL) and integral measurements.
About the SPIRE collaboration for spent fuel characterization.
LGI Consulting:
• Communication• Dissemination of results• Digital strategy and presence• Event management
● Energy dependence of neutron cross-section is a unique signature
● Attenuation of the neutron flux is linked to 239Pu content
Rossa R., et al. A new approach for the application of the Self-Interrogation Neutron Resonance Densitometry to spent fuel verifications. Symposium on International Safeguards, Linking Strategy, Implementation and People, Vienna, Austria, IAEA, 2014, p. 270
● How: fission chambers in the guide tubes of a PWR fuel assembly
● What: neutron flux in specific energy regions
● 239Pu fission chamber + Gd & Cd filters to select the 0.3 eV region
● 238U fission chamber to measure the fast neutron flux
Rossa R., et al. Optimization of the filters thickness for the SINRD technique applied to spent fuel verification.INMM 55th Annual Meeting Proceedings, Atlanta, GA, United States, 20-24 July 2014
● Air and polyethylene to ensure the best conditions
● Avoid moderation within the fuel assembly
● Definition of the SINRD signature Rossa R., et al. Investigation of the Self-Interrogation Neutron Resonance Densitometry applied to spent fuel using Monte Carlo simulations.Annals of Nuclear Energy 75 (2015) 176–183
D. Rochman, A.J. Koning and D. da Cruz, ``Propagation of 235,236,238U and 239Pu nuclear data uncertainties for a typical PWR fuel element'', Nuclear Technology 179, no. 3, 323-338 (2012).
Phenomenological data for simulation
• Nuclear simulation codes (MCNP, EASY, etc.) do not contain all physics of particle interaction (cross section, angular/energy emitted spectra.) but read from nuclear data tables
• Knowledge of nuclear interactions derived from careful and expensive experiments + sophisticated modelling
• Decades of effort has resulted in relatively reliable information for simulation of LWRs operation (not dismantlement) and other specific applications – using tiny fraction of nuclides/reactions
• Simulation of advanced reactors, geological storage will require substantially greater library with much more detail than simple σ(E)
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EASY-II roadmap1. FISPACT-2007+ & EAF-2010 in EAF format processed by SAFEPAQ-II ☑ 08/2010
2. FISPACT-II(11) & EAF-2010 in EAF format processed by SAFEPAQ-II ☑ 01/2011
3. FISPACT-II(11) & EAF-2010 + CALENDF PT’s ssf method, ENDF’s format and processing framework ☑ 09/2011
4. EASY-II(12) = FISPACT-II(12) & EAF’s and TENDL-2011 ENDF’s libraries processed by NJOY, PREPRO & CALENDF ☑ 03/2012
5. EASY-II(13) = FISPACT-II & EAF’s and TENDL’s V&V libraries processed by NJOY, PREPRO & CALENDF ☑ 06/2013
• The Total Monte Carlo (TMC) methodology uses direct feedback from simulation to physical inputs
• TMC provides truly remarkable uncertainty analysis based upon simulation outputs – where legacy provides little/none
• TMC is as good as the simulation capability. The marriage of TENDL with EASY provides the most robust methodology
By bringing the disjoint nuclear data links, from evaluation to application, into a technologically-driven closed system we can provide complete, robust data
superior to any legacy system
Current focus @ Uppsala
• Improved calibration of TENDL using both integral and differential data.
• UQ for fast and thermal reactor systems.
• Angular distributions– Uncertainties not well
evaluated 50
Measurement of independent fission yields in thermal and fast neutron spectra – IGISOL, Finland
To be done @ Uppsala
• Marriage between TALYS + GEF. – GEF is the best currently available nuclear fission model. GEF
will be the basis of future nuclear data evaluations by NEA/OECD within JEFF
• Calculate the decay heat in and burn-up of the fuel assemblies and estimate and minimize its uncertainties using the TMC methodology. Benchmark the results against both differential measurement (e.g. IGISOL) and integral measurement.
• Calculate criticality margins and associated uncertainties for both fuel storage and in deep repositories. Benchmark the results against both differential measurement (e.g. IGISOL) and integral measurement.
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Relations
UU, NRG, CCFE, (PSI)
TENDL,TMC, EASYII
SPIRE collaboration
Simulations and experiments for repositories
Nuclear data, code systems and U.Q.
Integral data for calibration and validation
E.ONs participation in SPIRE
• Uncertainties in neutron/ gamma dose and uncertainties of the decay heat are important factors for the safe and economical storage of spent nuclear fuel in interim storage facilities and in geological repositories.
• The uncertainties may be comparatively small on a fuel assembly basis but for the storage of hundreds and thousands of fuel assemblies the financial consequences are significant.
E.ONs participation in SPIRE:Motivation
• Minimization of the source term uncertainties, i.e:
• Accomplish source term uncertainty calculations for spent nuclear fuel given the latest covariance data from data libraries
• Second major aim is to adapt the latest technology from non-proliferation surveillance in order to significantly reduce the measurement uncertainties of the gamma and neutron dose and of the decay heat
E.ONs participation in SPIRE:Objective
• Path 1: Identification of gaps in the knowledge of the microscopic cross sections and nuclide yields and to fill these gaps with more data in order to reduce the microscopic data uncertainties.
• Path 2: Development of measurement equipment which can cut the current measurement uncertainties at least by a factor of two.
• These two routes are mutually reinforcing because• the better the measurements are, the better the knowledge gaps on the
microscopic data can be identified, and• the smaller the microscopic uncertainties are, the easier it is to design high
precision measurements.
E.ONs participation in SPIRE:Proposed approach
No Description1 Methodology development and demonstration of taking into account the
following uncertainties in source term calculation:- cross section uncertainties- nuclide yield uncertainties- power history uncertainties (i.e. burnup)- fuel assembly manufacturing tolerances
2 Improvement of measurement devices for decay heat, neutron & gamma dose and reduction of current uncertainties by at least a factor of two.
3 Measurement campaign at CLAB and/ or at NPP site to increase data base. Sample of fuel assemblies should cover shutdown cooling period between about 3 and 20 years. Repetitive measurement of homologous fuel assemblies with same, nominal burnup.
4 Evaluation of results of uncertainty analysis & measurement uncertainties to derive a final uncertainty of decay heat, n- and gamma dose.
E.ONs participation in SPIRE:Proposed focus
• Test and use the developed measurement equipment in one of its Nuclear Power Plants in order to determine the source terms on a suitable set of selected fuel assemblies
• Provide reactor power histories of selected fuel assemblies as input for theoretical calculations.
• This on-site measurement has the advantage that both fuel assemblies with a relatively short time after discharge and homologous pairs of fuel assemblies can be included into the database.
E.ONs participation in SPIRE:Potential contribution
SKB / DoE / Euratom / UU Collaboration
• Research effort to determine the capability of non-destructive assay (NDA) techniques for spent fuel.
• Partial defect detection• Heat quantification• Assembly operational parameters (IE, BU, CT)• Pu mass• Reactivity determination
• To meet the combined needs of … • the safeguards community• the operator (SKB)
SKB / DoE / Euratom / UU Collaboration
Experimental signatures measured at Clab from…
• Spectral resolved gammas (HPGe and LaBr3).
• Time correlated neutrons (Differential Die-away Self Interrogation).
• Time-varying and continuous active neutron interrogation (Differential Die-Away).
• An approximation of Californium Interrogation Prompt Neutron (CIPN).
• Total neutron and total gamma fluxes (FORK Detector).
• Total decay heat (assembly length calorimeter).
• Possibly also Cerenkov light emission (Digital Cerenkov Viewing Device).
“On-going” spent fuel measurements at Clab:• Calorimetric measurements of decay heat.• Passive gamma measurements (HPGe).
Measurements currently planned to be performed at Clab:• Passive neutron measurements (FORK).• Passive gamma measurements (HPGe, LaBr3).• Differential die-away self interrogation measurements
• Partial defect detection• Heat quantification• Assembly operational parameters (IE, BU, CT)• Pu mass• Reactivity determination
SKB / DoE / Euratom / UU Collaboration
Determined using...• the experimental signatures• data mining (“analytic solver”)
S. Tobin et al, “Experimental and Analytical Plans for the Non-destructive Assay System of the Swedish Encapsulation and Repository Facilities”,IAEA Symposium on International Safeguards: Linking Strategy, Implementation and People. 2014: https://www.iaea.org/safeguards/symposium/2014/home/eproceedings/sg2014-papers/000238.pdf