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DOE OSTI2012

Full-Scale Cask Testing and Public Acceptance of Spent Nuclear Fuel Shipments - 12254

Full-scale physical testing of spent fuel shipping casks has been proposed by the National Academy of Sciences (NAS) 2006 report on spent nuclear fuel transportation, and by the Presidential Blue Ribbon Commission (BRC) on America's Nuclear Future 2011 draft report. The U.S. Nuclear Regulatory Commission (NRC) in 2005 proposed full-scale testing of a rail cask, and considered 'regulatory limits' testing of both rail and truck casks (SRM SECY-05-0051). The recent U.S. Department of Energy (DOE) cancellation of the Yucca Mountain project, NRC evaluation of extended spent fuel storage (possibly beyond 60-120 years) before transportation, nuclear industry adoption of very large dual-purpose canisters for spent fuel storage and transport, and the deliberations of the BRC, will fundamentally change assumptions about the future spent fuel transportation system, and reopen the debate over shipping cask performance in severe accidents and acts of sabotage. This paper examines possible approaches to full-scale testing for enhancing public confidence in risk analyses, perception of risk, and acceptance of spent fuel shipments. The paper reviews the literature on public perception of spent nuclear fuel and nuclear waste transportation risks. We review and summarize opinion surveys sponsored by the State of Nevada over the past two decades, which show consistent patterns of concern among Nevada residents about health and safety impacts, and socioeconomic impacts such as reduced property values along likely transportation routes. We also review and summarize the large body of public opinion survey research on transportation concerns at regional and national levels. The paper reviews three past cask testing programs, the way in which these cask testing program results were portrayed in films and videos, and examines public and official responses to these three programs: the 1970's impact and fire testing of spent fuel truck casks at Sandia National Laboratories, the 1980's regulatory and demonstration testing of MAGNOX fuel flasks in the United Kingdom (the CEGB 'Operation Smash Hit' tests), and the 1980's regulatory drop and fire tests conducted on the TRUPACT II containers used for transuranic waste shipments to the Waste Isolation Pilot Plant in New Mexico. The primary focus of the paper is a detailed evaluation of the cask testing programs proposed by the NRC in its decision implementing staff recommendations based on the Package Performance Study, and by the State of Nevada recommendations based on previous work by Audin, Resnikoff, Dilger, Halstead, and Greiner. The NRC approach is based on demonstration impact testing (locomotive strike) of a large rail cask, either the TAD cask proposed by DOE for spent fuel shipments to Yucca Mountain, or a similar currently licensed dual-purpose cask. The NRC program might also be expanded to include fire testing of a legal-weight truck cask. The Nevada approach calls for a minimum of two tests: regulatory testing (impact, fire, puncture, immersion) of a rail cask, and extra-regulatory fire testing of a legal-weight truck cask, based on the cask performance modeling work by Greiner. The paper concludes with a discussion of key procedural elements - test costs and funding sources, development of testing protocols, selection of testing facilities, and test peer review - and various methods of communicating the test results to a broad range of stakeholder audiences. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2012

The EU Approach for Responsible and Safe Management of Spent Fuel and Radioactive Waste - 12118

In July 2011 legislation on responsible and safe management of spent fuel and radioactive waste was adopted in the European Union (EU). It aims at ensuring a high level of safety, avoiding undue burdens on future generations and enhancing transparency. EU Member States are responsible for the management of their spent fuel and/or radioactive waste. Each Member State remains free to define its fuel cycle policy. The spent fuel can be regarded either as a valuable resource that may be reprocessed or as radioactive waste that is destined for direct disposal. Whatever option is chosen, the disposal of high level waste, separated at reprocessing, or of spent fuel regarded as waste should be considered. The storage of radioactive waste, including long-term storage, is an interim solution, but not an alternative to disposal. To this end, each Member State has to establish, maintain and implement national policy, framework and programme for management of spent fuel and/or radioactive waste in the long term. Member States will invite international peer reviews to ensure that high safety standards are achieved. The EU approach is anchored in internationally endorsed principles and requirements of the IAEA safety standards and the Joint Convention and in this context makes them legally binding and enforceable in the EU. The EU approach of regulating the management of spent fuel and radioactive waste is anchored in the competence of the national regulatory authorities and in the internationally endorsed principles and requirements of the IAEA Safety Standards and the Joint Convention. Member States have to report to the Commission on the implementation of Directive 2011/70/Euratom for the first time by 23 August 2015, and every 3 years thereafter, taking advantage of the review and reporting under the Joint Convention. On the basis of the Member States' reports, the Commission will submit to the European Parliament and the Council a report on progress made and an inventory of radioactive waste and spent fuel present in the EU territory and the future prospects. Directive 2011/70/Euratom is a logical next step after the Council Directive 2009/71/Euratom on the nuclear safety of nuclear installations. The EU is the first major regional actor providing a binding legal framework on nuclear safety and on responsible and safe management of spent fuel and radioactive waste, and thus is a real model to progress spent fuel and waste management in a safe and responsible manner. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2018

Development of Neutron Energy Spectral Signatures for Passive Monitoring of Spent Nuclear Fuels in Dry Cask Storage

Demand for spent nuclear fuel dry casks as an interim storage solution has increased globally and the IAEA has expressed a need for robust safeguards and verification technologies for ensuring the continuity of knowledge and the integrity of radioactive materials inside spent fuel casks. Existing research has been focusing on “fingerprinting” casks based on count rate statistics to represent radiation emission signatures. The current research aims to expand to include neutron energy spectral information as part of the fuel characteristics. First, spent fuel composition data are taken from the Next Generation Safeguards Initiative Spent Fuel Libraries, representative for Westinghouse 17 x 17 PWR assemblies. The ORIGEN-S code then calculates the spontaneous fission and (α,n) emissions for individual fuel rods, followed by detailed MCNP simulations of neutrons transported through the fuel assemblies. A comprehensive database of neutron energy spectral profiles is to be constructed, with different enrichment, burn-up, and cooling time conditions. The end goal is to utilize the computational spent fuel library, predictive algorithm, and a pressurized 4 He scintillator to verify the spent fuel assemblies inside a cask. This work identifies neutron spectral signatures that correlate with the cooling time of spent fuel. Both the total and relative contributions from spontaneous fission and (α,n) change noticeably with respect to cooling time, due to the relatively short half-life (18 years) of the major neutron source 244 Cm. Identification of this and other neutron spectral signatures allows the characterization of spent nuclear fuels in dry cask storage.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Simulation of Spent Nuclear Fuel loading into a Final Waste Repository

Previous work towards studying repository loading have used: - spent nuclear fuel (SNF) with an average burnup composition; - a lumped capacitance thermal model for calculating temperature in a Cyclus repository model. The goal of this work is to improve on the repository models and use U.S. historical SNF inventory data in simulations to more accurately study the loading of a waste repository. These goals will be achieved by: - using UNF-ST-and-DARDS Unified Database (UDB) that has historic assembly-specific data (e.g, isotopic composition, heat) in Cyclus simulations; - implementing a more accurate thermal model within a Cyclus repository model. Objectives: - Create a Cyclus spent fuel conditioning model that packages spent fuel bundles into packages which have user-defined properties. - Create a Cyclus medium-fidelity repository model that gives accurate time and spatial dependent temperature values and loads the repository based on a user-selected loading strategy. Cyclus is an agent-based extensible framework for modeling flow of material through user-defined nuclear fuel cycles. In Cyclus, each facility in the fuel cycle is modeled individually and the facilities interact with one another as independent agents. The spent fuel conditioning model accepts spent fuel bundles and puts them into a cylindrical waste package. In the spent fuel conditioning model, the user can define variables: For each layer, radius, thermal conductivity, thermal diffusivity; For each package, Number of spent fuel bundles, Radius and height. The waste repository model accepts waste packages and emplaces them into available positions within the waste repository based on a thermal criteria. The thermal criteria is a temperature limit at the interface between the waste package surface and the host geology, that is set based on the repository's host geology. In the waste repository model, the user can define the variables: Capacity, Distance between waste packages, Distance between drifts, Repository host geology, Loading Strategy. After the addition of new waste packages at each time step, the waste repository model recalculates the temperature at each location in the repository. If the addition of this new package causes its temperature to exceed the thermal limit, it will be placed back into the buffer. A thermal model that relies on a transient 'outside' model and quasi-steady-state 'inside' model is used to accurately determine the temperature in the repository. The 'outside' model assumes a homogenous medium with the Engineered Barrier System (EBS) replaced by the geologic medium. Temperature solutions for the central waste package, adjacent point and line sources are superimposed to calculate the temperature at specific points in the repository. The central drift consists of one finite line source which represents the central waste package. The central drift also consists of point sources that represent neighboring waste packages in the central drift. The neighboring drifts are represented by infinite line sources. The 'inside' model is considered to be at a quasi-steady-state condition because EBS has a relatively low thermal mass compared to the infinite geologic medium. The steady state calculation is performed at each time step with the heat source and interface temperature as boundary conditions. Future Work: Run Cyclus simulations with U.S. historical SNF inventory data, the spent fuel conditioning and repository models to study how waste package acceptance strategies impact repository loading.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

In-Field Performance Testing of the Fork Detector for Quantitative Spent Fuel Verification

Expanding spent fuel dry storage activities worldwide are increasing demands on safeguards authorities that perform inspections. The European Atomic Energy Community (EURATOM) and the International Atomic Energy Agency (IAEA) require measurements to verify declarations when spent fuel is transferred to difficult-to-access locations, such as dry storage casks and the repositories planned in Finland and Sweden. EURATOM makes routine use of the Fork detector to obtain gross gamma and total neutron measurements during spent fuel inspections. Data analysis is performed by modules in the integrated Review and Analysis Program (iRAP) software, developed jointly by EURATOM and the IAEA. Under the framework of the US Department of Energy–EURATOM cooperation agreement, a module for automated Fork detector data analysis has been developed by Oak Ridge National Laboratory (ORNL) using the ORIGEN code from the SCALE code system and implemented in iRAP. EURATOM and ORNL recently performed measurements on 30 spent fuel assemblies at the Swedish Central Interim Storage Facility for Spent Nuclear Fuel (Clab), operated by the Swedish Nuclear Fuel and Waste Management Company (SKB). The measured assemblies represent a broad range of fuel characteristics. Neutron count rates for 15 measured pressurized water reactor assemblies are predicted with an average relative standard deviation of 4.6%, and gamma signals are predicted on average within 2.6% of the measurement. The 15 measured boiling water reactor assemblies exhibit slightly larger deviations of 5.2% for the gamma signals and 5.7% for the neutron count rates, compared to measurements. These findings suggest that with improved analysis of the measurement data, existing instruments can provide increased verification of operator declarations of the spent fuel and thereby also provide greater ability to confirm integrity of an assembly. These results support the application of the Fork detector as a fully quantitative spent fuel verification technique.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Safeguards Approaches for Very Long-Term Storage of Spent Fuel

Many States are considering interim very long-term storage (VLTS) strategies for spent fuel storage while final disposition options are developed. Under the VLTS approach, spent fuel will be stored for periods that could exceed 50 to 100 years. The long-term storage of spent fuel — and the increasing stockpile of plutonium contained in this spent fuel — exacerbate the international safeguards and verification needs for spent fuel. In particular, significant delays in any verification of the contents of spent fuel assemblies prior to disposition may cause even further delays in the final disposition (or reprocessing) of spent fuel. As a result, the resources required by the International Atomic Energy Agency (IAEA) to safeguard the increasing stockpile of spent fuel will increase. Safeguarding these interim storage facilities and subsequently, the transportation of long-term stored fuel, will place additional demands on IAEA resources if the current approach, which relies heavily upon inspectors being present at the facility, is to be continued. Current IAEA approaches to safeguarding above ground dry storage sites do not take into account the challenges associated with longer-term storage and are also very dependent on inspector presence. The IAEA is moving toward remote and unattended monitoring systems for safeguards and these systems should be utilized at VLTS facilities as much as possible. The research and development of various monitoring systems should be continued.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2011

A Second Look at Neutron Resonance Transmission Analysis as a Spent Fuel NDA Technique

Many different nondestructive analysis techniques are currently being investigated as a part of the United States Department of Energy's Next Generation Safeguards Initiative (NGSI) seeking methods to quantify plutonium in spent fuel. Neutron Resonance Transmission Analysis (NRTA) is one of these techniques. Having first been explored in the mid-1970s for the analysis of individual spent-fuel pins a second look, using advanced simulation and modeling methods, is now underway to investigate the suitability of the NRTA technique for assaying complete spent nuclear fuel assemblies. The technique is similar to neutron time-of-flight methods used for cross-section determinations but operates over only the narrow 0.1-20 eV range where strong, distinguishable resonances exist for both the plutonium (239, 240, 241,242Pu) and uranium (235,236,238U) isotopes of interest in spent fuel. Additionally, in this energy range resonances exists for six important fission products (99Tc, 103Rh, 131Xe, 133Cs, 145Nd, and 152Sm) which provide additional information to support spent fuel plutonium assay determinations. Initial modeling shows excellent agreement with previously published experimental data for measurements of individual spent-fuel pins where plutonium assays were demonstrated to have a precision of 2-4%. Within the simulation and modeling analyses of this project scoping studies have explored fourteen different aspects of the technique including the neutron source, drift tube configurations, and gross neutron transmission as well as the impacts of fuel burn up, cooling time, and fission-product interferences. These results show that NRTA may be a very capable experimental technique for spent-fuel assay measurements. The results suggest sufficient transmission strength and signal differentiability is possible for assays through up to 8 pins. For an 8-pin assay (looking at an assembly diagonally), 64% of the pins in a typical 17 ? 17 array of a pressurized water reactor fuel assembly can be part of a complete transmission assay measurement with high precision. Analysis of rows with up to 12 pins may also be feasible but with diminished precision. Preliminary data analysis of an NRTA simulation has demonstrated the simplicity of the technique.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2015

Applying fast calorimetry analysis on spent fuel calorimeter - 15620

Swedish Nuclear Fuel and Waste Management Company (SKB) is responsible for the back-end of the Swedish nuclear industry. This includes transport of spent fuel by boat from the power plants to the Central Interim Storage Facility for Spent Fuel (Clab), and in the future, encapsulation of the spent fuel and further transportation to a geological repository. In order to ensure safety for the planned geological repository, the decay heat has to be well known. For this purpose the SKB has performed calorimetric measurements on over 200 pressurized water reactor (PWR) and boiling water reactor (BWR) assemblies since 2003. The results from earlier analyses have been used to improve computer codes predicting heat as well as to acquire deeper knowledge about the fuel inventory. The work complements research being done as part of the Next Generation of Safeguards Initiative - Spent Fuel project, whose purpose is to develop and test nondestructive assay (NDA) technologies to improve safeguards measurements of spent nuclear fuel. The project is working to accomplish multiple goals, one of which is to determine the heat content of spent fuel assemblies. To shorten the measurement time, the temperature increase method was used instead of the equilibrium temperature method. In the temperature increase method, the temperature inside of the calorimeter is lowered below the surrounding water temperature before the measurement, and the temperature is allowed to increase due to the higher surrounding water temperature during the measurement. The rate of the temperature increase when the temperature is the same as the surrounding temperature is then compared to a calibration curve where an electric heater has been used in the same way. The price for the higher throughput is less accuracy and dependency of fuel geometry since the heat flow will be different for different fuel types, especially between PWR and BWR assemblies. Recently at Los Alamos National Laboratory, more sophisticated analysis methods have been considered for the use of calorimetry for treaty verification. These methods aim to predict the equilibrium temperature based on early data and therefore be able to shorten the measurement time while maintaining good accuracy and eliminate the dependency on fuel geometry. Even though the techniques show promise for future calorimetry of spent fuel, certain design features of the current system have been identified to interfere with the accuracy of equilibrium measurements in general such as changing pool temperatures and heat leakage during calibration. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Accident analysis of spent fuel storage and transportation cask with Gd based neutron absorber

The shortage of nuclear spent fuel storage pool has been issued. The spent fuel storage and transportation cask is one of the major alternatives for storing spent fuel. The use of advanced neutron absorber is important for effective storage of spent fuel in spent fuel cask. Gadolinium based neutron absorber is chosen since it has high neutron absorption cross section in thermal neutron region compared to boron and it does not produce any gas element when it reacts with a neutron. The major requirement for neutron absorbing materials in the spent fuel cask system is to maintain subcriticality in normal, off-normal and accident conditions. In this study, the criticality calculations under accident conditions are conducted to ensure the safety of spent fuel cask. Major accident conditions are flood and cask tip-over. The full and partial flooded condition is analyzed and 100% flooded condition has the highest reactivity. The water density is selected as 1 g/cm{sup 3} because k{sub eff} is highest at this condition. When the cask tip-over, the flux trap reduces. In this case, a huge increase in reactivity of 800 pcm takes place. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2014

Total Integrated Dose to Spent Fuel Pool Level Instrument Probes and Refueling Floor Personnel Dose Rates Meeting NRC Order EA-12-051 for a Typical Boiling Water Reactor - Paper 22

In reviewing the events after the March 11, 2011 Fukushima accident, the Nuclear Regulatory Commission (NRC) issued Order EA-12-051, Reference 1, requiring new reliable spent fuel pool (SFP) instrumentation for pool level determination under beyond design basis event conditions. In 2013, URS prepared a calculation for a typical Boiling Water Reactor to determine the Total Integrated Dose (TID) potentially to be imparted to radiation sensitive underwater probes (one primary and one backup) planned for meeting this Order. Additionally, URS calculated the radiological dose rates to an individual standing on the refueling floor for various SFP fill levels. NEI 12-02, Reference 2, defines the levels at which SFP condition identification is necessary as the following: Level 1 - level that is adequate to support operation of the normal fuel pool cooling system. Level 1 represents the higher of the following two points: - The level at which reliable suction loss occurs due to uncovering of the coolant inlet pipe, weir or vacuum breaker (depending on the design), or - The level at which the water height, assuming saturated conditions, above the centerline of the cooling pump suction provides the required net positive suction head specified by the pump manufacturer or engineering analysis. Level 2 - level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck. Level 2 is based on either of the following: - 10 feet (+/- 1 foot) above the highest point of any fuel rack seated in the spent fuel pools, or - A designated level that provides adequate radiation shielding to maintain personnel radiological dose levels within acceptable limits while performing local operations in the vicinity of the pool. This level shall be based on either plant-specific or appropriate generic shielding calculations, considering the emergency conditions that may apply at the time and the scope of necessary local operations, including installation of portable SFP instrument channel components. Level 3 - level where fuel remains covered and actions to implement make-up water addition should no longer be deferred. Level 3 corresponds nominally (i.e. +/- 1 foot) to the highest point of any fuel rack seated in the spent fuel pool. The URS calculation did not define these levels or determine instrumentation setpoints. However, it determined refueling floor dose rates that could be used as a tool to determine the appropriate water depths for these levels. In determining the TID for the SFP probes and the radiological dose rates to the individual, the following radioactive sources were considered: 1. Spent Nuclear Fuel (SNF) stored in the spent fuel storage racks, 2. Local Power Range Monitors (LPRMs) stored above the top of the racks on hangers on the SFP walls 3. Radioactive material laden filters stored on platforms on top of the racks, and 4. Control Rod Blades (CRBs) stored above the top of the racks on hangers on the SFP walls. The TID for the SFP probes includes both the dose from normal operation (pool full of water) for the remaining plant life (an assumed 40 years, conservatively including an allowance for lifetime extension SFP loading) and, consistent with the NEI 12-02 guidelines, Reference 2, the assumed drop in elevation of the pool water level to the top of the spent fuel racks for a 6 hour period. The acceptable TID for the radiation-sensitive probe as defined by the manufacturer was 2x10{sup 9} rad. The individual on the pool deck was assumed to be standing on the edge of the pool with an unobstructed view of its contents, as a worst case scenario. Dose rates to this individual determined from direct shine from the same 4 sources were determined for SFP water levels varying from 1 ft above the SNF racks to 21 ft above the racks in one foot increments. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Improved evaluation of safeguards parameters from spent fuel measurements with the Differential Die-Away (DDA) instrument

The Differential Die-Away (DDA) technique is a highly sensitive non-destructive assay method for characterizing and detecting the presence of fissile material within an item of interest. DDA utilizes a series of pulses from a neutron generator (NG) to actively interrogate an item of interest. The die-away time of the neutron population induced by this active interrogation and the integral of the total differential die-away signal can be used to characterize items such as nuclear waste drums and spent nuclear fuel assemblies. In this work, Los Alamos National Laboratory (LANL) conceptualized, designed, and fabricated a DDA instrument that was deployed for field test measurements at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab) in Oskarshamn, Sweden. The instrument performed multiple static measurements at fixed locations and dynamic axial scans of 15 pressurized water reactor (PWR) and 10 boiling water reactor (BWR) spent fuel assemblies, collecting both passive and active measurement data. The static assays of the assemblies measured the differential die-away signal, die-away time, and total passive neutron emission rate to create calibration curves for the evaluation of assembly multiplication, burnup, initial enrichment, effective fissile mass, and total elemental plutonium mass. Each calibration curve was optimized by minimizing the relative root mean square error (RRMSE) of assembly assay results compared to declared assembly parameters. The same quantities were also measured with the axial scans, and the resulting data were applied in two ways: (1) in the creation of calibration curves to improve evaluation of the same safeguards parameters as static assays, and (2) for comparison to simulation. In most cases, across both PWR and BWR assemblies, axial scan data improved the estimation of the above parameters, quantified by decreasing the calibration curve RRMSE. These axial scan results demonstrate the ability of the DDA instrument and analysis method to characterize spent PWR and BWR fuel as well as, or better than, a static assay of the same assembly. Furthermore, the DDA instrument’s unique ability to obtain both active and passive data in a single, axial scan of an entire spent fuel assembly represents a more efficient and accurate way of assaying spent fuel for verification purposes. These results represent a significant advancement for characterizing spent nuclear fuel compared to current technologies.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2019

Dry Cask Shipments of ATR Spent Nuclear Fuel at INL

The High Load Charger (HLC) cask will be replacing the Advanced Test Reactor (ATR) spent fuel transfer cask that has been used at Idaho National Laboratory (INL) to wet-transfer spent ATR fuel elements from ATR to Idaho Nuclear Technology and Engineering Center (INTEC) in order to support future plans at both facilities. INTEC intends to discontinue use of the CPP-666 storage pools in compliance with the Idaho Settlement Agreement and the basins will no longer be available to store spent ATR fuel elements. In order to support the core-internals change out in 2021 for ATR, spent ATR fuel elements must be removed from the canal and shipped to INTEC. INL contractors will use the HLC cask for safe, dry transfer of spent Advanced Test Reactor (ATR) fuel elements between the ATR Complex managed by Battelle Energy Alliance, LLC and the Idaho Nuclear Technology and Engineering Center (INTEC) managed by Fluor Idaho, LLC. The HLC cask (also known as the High Load Charging Cask or High Load Charger) is a lead-shielded, stainless steel transfer container originally designed in 1954 by the Idaho Nuclear Corporation to meet the radiation and contamination controls requirements of the time for shipment of spent nuclear fuel. The cask has been modified over the years and is currently used for wet-to-dry transfers at INTEC. It is approved only for out-of-commerce transfers. The cask will arrive at the ATR where eight spent ATR fuel elements will be loaded into two, newly-designed buckets. Each bucket is a quarter section, right-cylinder shape that has four compartments. The cask will be lifted out of the ATR canal allowing most of the water to drain from the cask. Processing the 12,500 kg cask will be completed in ATR and then transported on a tractor trailer to INTEC where the loaded buckets will be removed from the cask. The loaded buckets will undergo a drying process before placing the four buckets into a canister for long-term storage in the Irradiated Fuel Storage Facility (IFSF) at INTEC. A new transportation plan that was submitted to, and approved by, the U.S. Department of Energy to provide the cask safety basis, which describes the hazard analysis and accident analysis for the cask during transport. Supporting accident analysis include criticality, thermal, dose, and drop analyses. Of particular interest for this cask was the drop analysis from which was derived the speed limit for the transport. The drop analysis addressed drops in air onto an unyielding surface and in water onto an energy-absorbing pad for the cask in the ATR facility. These analyses are typically applied as the bounding drop scenarios during transport; however, this resulted in a safety basis speed limit of 12.9 kph. In order to more accurately represent the in-transit accident scenario, LS-DYNA software was procured and the model was updated to more closely represent the potential drop heights and impact surfaces of asphalt and soil for the cask while in transit. This analysis resulted in a safety basis speed limit of 48 kph. This safety basis speed limit is more practical to implement and control.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Validation of ORIGEN for VVER-440 Spent Fuel with Application to Fork Detector Safeguards Measurements

The US Department of Energy National Nuclear Security Administration and the European Atomic Energy Community (Euratom) are collaborating with the Radiation and Nuclear Safety Authority in Finland to assess spent fuel verification methods for potential application to measurements at the planned spent fuel encapsulation plant and geological repository prior to final disposal in Finland. The fork detector (FDET) used to measure the neutron and gamma radiation from a given spent fuel assembly is an existing technology widely used by the International Atomic Energy Agency (IAEA) and Euratom for the safeguards verification of spent fuel operator declaration. Recently, an FDET data analysis software, referred to as the ORIGEN module, was developed and incorporated into the IAEA/Euratom Integrated Review and Analysis Program. This module uses the ORIGEN burnup code to calculate the nuclide concentrations and the neutron and gamma source terms in the spent fuel based on operator declarations. Then, to predict the expected FDET signals for a given fuel assembly, the module combines the source terms calculated by ORIGEN with FDET response functions that were pre-generated using MCNP models. The inspector can compare the calculated signals to the measured values in real time to identify possible anomalies in fuel assembly’s operator declaration, integrity, or measurement. In this work, the capability of the ORIGEN module was extended from application to typical light water reactor fuels to use for VVER-440 assembly types. The accuracy of ORIGEN for this type of fuel assembly calculations was assessed by comparing calculated nuclide concentrations against destructive assay measurements for VVER-440 spent fuel. The performance of the ORIGEN module for FDET safeguards verification in routine inspections was assessed using FDET measurement data for 13 spent VVER-440 assemblies that were measured at the Loviisa Nuclear Power Plant in Finland, by comparing the calculated signals to the measured quantities. The results show that the calculated FDET neutron and gamma detector signals are generally within 12% of the measurements except for one assembly. Thus, these results are applicable to future safeguards verification in the planned Finland encapsulation plant, such as for passive neutron albedo reactivity safeguards measurements.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

DOE OSTI2012

AP1000{sup R} nuclear power plant safety overview for spent fuel cooling

The AP1000{sup R} plant is an 1100-MWe class pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and costs. The AP1000 design uses passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems such as AC power, component cooling water, service water or HVAC. Furthermore, these passive features 'fail safe' during a non-LOCA event such that DC power and instrumentation are not required. The AP1000 also has simple, active, defense-in-depth systems to support normal plant operations. These active systems provide the first level of defense against more probable events and they provide investment protection, reduce the demands on the passive features and support the probabilistic risk assessment. The AP1000 passive safety approach allows the plant to achieve and maintain safe shutdown in case of an accident for 72 hours without operator action, meeting the expectations provided in the U.S. Utility Requirement Document and the European Utility Requirements for passive plants. Limited operator actions are required to maintain safe conditions in the spent fuel pool via passive means. In line with the AP1000 approach to safety described above, the AP1000 plant design features multiple, diverse lines of defense to ensure spent fuel cooling can be maintained for design-basis events and beyond design-basis accidents. During normal and abnormal conditions, defense-in-depth and other systems provide highly reliable spent fuel pool cooling. They rely on off-site AC power or the on-site standby diesel generators. For unlikely design basis events with an extended loss of AC power (i.e., station blackout) or loss of heat sink or both, spent fuel cooling can still be provided indefinitely: - Passive systems, requiring minimal or no operator actions, are sufficient for at least 72 hours under all possible pool heat load conditions. - After 3 days, several different means are provided to continue spent fuel cooling using installed plant equipment as well as off-site equipment with built-in connections. Even for beyond design basis accidents with postulated pool damage and multiple failures in the passive safety-related systems and in the defense-in-depth active systems, the AP1000 multiple spent fuel pool spray and fill systems provide additional lines of defense to prevent spent fuel damage. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

A Non-Proliferating Fuel Cycle: No Enrichment, Reprocessing or Accessible Spent Fuel - 12375

Current fuel cycles offer a number of opportunities for access to plutonium, opportunities to create highly enriched uranium and access highly radioactive wastes to create nuclear weapons and 'dirty' bombs. The non-proliferating fuel cycle however eliminates or reduces such opportunities and access by eliminating the mining, milling and enrichment of uranium. The non-proliferating fuel cycle also reduces the production of plutonium per unit of energy created, eliminates reprocessing and the separation of plutonium from the spent fuel and the creation of a stream of high-level waste. It further simplifies the search for land based deep geologic repositories and interim storage sites for spent fuel in the USA by disposing of the spent fuel in deep sub-seabed sediments after storing the spent fuel at U.S. Navy Nuclear Shipyards that have the space and all of the necessary equipment and security already in place. The non-proliferating fuel cycle also reduces transportation risks by utilizing barges for the collection of spent fuel and transport to the Navy shipyards and specially designed ships to take the spent fuel to designated disposal sites at sea and to dispose of them there in deep sub-seabed sediments. Disposal in the sub-seabed sediments practically eliminates human intrusion. Potential disposal sites include Great Meteor East and Southern Nares Abyssal Plain. Such sites then could easily become international disposal sites since they occur in the open ocean. It also reduces the level of human exposure in case of failure because of the large physical and chemical dilution and the elimination of a major pathway to man-seawater is not potable. Of course, the recovery of uranium from sea water and the disposal of spent fuel in sub-seabed sediments must be proven on an industrial scale. All other technologies are already operating on an industrial scale. If externalities, such as reduced terrorist threats, environmental damage (including embedded emissions), long term care, reduced access to 'dirty' bomb materials, the social and political costs of siting new facilities and the psychological impact of no solution to the nuclear waste problem, were taken into account, the costs would be far lower than those of the present fuel cycle. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Investigation of Imaging Spent Nuclear Fuel Dry Casks using Cosmic Ray Muons

Recently, the use of cosmic ray muons for cargo scanning applications has been demonstrated and muons have been shown to have the potential to allow for nondestructive assessment of nuclear material accountancy. Relativistic muons have the ability to penetrate dense materials and monitoring the subsequent scattering of muons can provide information about the structure and composition of materials stored in cargo containers. The suitability of cosmic muons for a number of monitoring and imaging applications has been investigated over the years and includes applications to archaeology, volcano imaging, material identification and medical diagnosis. Novel applications of cosmic ray muons have been proposed for examination of comatose patients towards bone density monitoring and also for determination of the molten nuclear fuel location in nuclear reactors having suffered from the effects of a severe accident similar to the one happened in Fukushima. There is a need to develop a capability to monitor nuclear waste and investigate whether the stored content agrees with facility declarations to allow proliferation detection and international treaty verification. Monitoring nuclear waste and controlling nuclear material at its source is one of the main strategies to minimize the risks of nuclear proliferation and reduce potential homeland threats. Since the early 1950's, when the first nuclear power plant began to produce electricity, approximately 65,000 metric tons of spent nuclear fuel have been generated, 25% of which are under dry storage conditions. After the spent nuclear fuel has been placed inside the dry cask, the cask is welded, not allowing for visual inspection. During the long storage life of the cask, intermediate handling and transportation activities take place which does create concerns regarding fuel integrity. Accidents and natural disasters also contribute to this concern. Previous work on spent nuclear fuel dry cask monitoring using muons has shown that the muon scattering distributions are adequately separated and a and a decision rule derived from Bayes' Theorem rendering the classification of different dry casks feasible. Muons present significant advantages over the existing imaging techniques such as the utilization of the passive nature of muons, the lack of radiological sources and consequently the absence of any artificial radiological dose. In the present work, a methodology is developed to demonstrate the applicability of muons for imaging of spent nuclear fuel dry casks. Purpose is to use muons to differentiate between spent nuclear fuel dry casks with different amounts of loading. Monte Carlo simulations of dry casks with different spent nuclear fuel loading are performed to simulate the passage of polyenergetic muons having energies in the range of 1-60 GeV through matter. A 'Muon Generator' is developed to generate muons that follow the actual sea level energy distribution. The 'Muon Generator' is validated against a series of actual spectrum measurements and then coupled with the Monte Carlo particle transport code GEANT4. The accumulated dataset of scattering angles were processed with the Point-of- the-Closest-Approach (PoCA) imaging algorithm and the obtained images demonstrate that an empty dry cask can be clearly distinguished from a half-loaded or a fully loaded one. The apparent separation of the images demonstrates that the muon scattering can be used as a feature for imaging spent nuclear fuel dry casks. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Dry Cask Shipments of ATR Spent Nuclear Fuel at INL - 19600

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

The Effect of Localized Pin Flux on Short-Term Gamma Intensities of Spent Fuel Assemblies

Nuclear reactor operation for electricity generation involves irradiation of a fuel assembly for an extended period of time until a certain burnup is achieved. To harness the full nuclear energy potential and flatten the flux around the core, fuel assemblies are shuffled to one or more different locations within the core. Post-irradiation practice typically involves storing the spent fuel assembly in a spent fuel pool. While in the spent fuel pool, aspects of the fuel assembly continue to evolve, particularly the isotopic concentration as nuclides decay. Numerous nondestructive assay (NDA) safeguards techniques exist and have been proposed for performing verification on spent fuel assemblies to ensure they have not been tampered with. Adigun et al. recently proposed measuring both the neutron and gamma signatures from spent fuel from discharge to final disposal, however, the research in only investigated the neutron part of this approach. In the conclusion, the authors caveated the need for the gamma results to make more certain conclusions about a spent fuel assembly, as a combination of the neutron and gamma NDA measurements will limit the range of possible initial enrichment and burnup combinations. This paper focuses on the production of the gamma source. It is important to understand the driving mechanisms behind the emissions signatures at shutdown and after long cooling time periods, to better determine the correlation between the measured passive gamma emission spectra and the gamma emissions per second of the assembly. The objective is to demonstrate how the effect of localized flux within a reactor affects the gamma emissions per second (or gamma intensity) present in an assembly at shutdown. This study examined the photons emitted from the nuclides present within the fuel assembly, and not the photons escaping from the fuel assembly. The latter point of transporting the photons present within the spent fuel assemblies to an appropriate detector will be addressed in a future paper. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗