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DOE OSTI2019

Guidance on Good Practice for Waste Management in Decommissioning - 19080

The World Nuclear Association (WNA) is an international organization that represents the global nuclear industry. Its mission is to promote a wider understanding of nuclear energy among key international influencers by producing authoritative information, developing common industry positions, and contributing to the energy debate. The Waste Management and Decommissioning Working Group of WNA monitors developments and shapes industry positions in radioactive waste management in nuclear decommissioning. The Working Group promotes accepted international principles and methods, including the sustainable management of natural resources, appropriate re-use and recycling of material, as well as safe and cost-efficient disposal of radioactive waste. In 2015, the WNA decided to identify international good practices in decommissioning and waste management and to appoint a group of international experts to produce a report bringing together this knowledge. The report is aimed at policy- and decision-makers, as well as nuclear professionals looking for guidance or facing decommissioning challenges related to waste management. The scope of the report covers commercial nuclear plants and nuclear research facilities, including legacy plants being decommissioned after normal operation. Although light water and graphite reactor types are mainly considered, the methods described can be applied to other nuclear facilities. The report is due to be published in February 2019. This paper summarizes its findings and recommendations. (authors)

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DOE OSTI2017

The Status of LILW Complex Disposal Facility Construction in Korea - 17123

In this paper, the experiences and directions of the construction of the LILW disposal facility in Korea are discussed. The LILW disposal facility is being constructed in Gyeongju-si, Korea. The disposal facility is officially called 'Wolsong Low and Intermediate Level Radioactive Waste Disposal Center (WLDC)'. It can dispose of 800,000 drums (hereafter, 'drums' means '200 liter drum equivalents' unless specified otherwise) of radioactive waste in a site of 2,060,000 m{sup 2}. The first stage of the LILW facility construction was completed in June 2014, and approved for use by Nuclear Safety and Security Commission (NSSC) in December 2014. The first stage of the LILW facility is a type of underground silo with disposal capacity of 35,200 m{sup 3} (100,000 drums). The first stage of the LILW facility consists of a construction tunnel, an operation tunnel, an access shaft and six silos, etc. Silo for final disposal is located 80∼130 m below sea level, 25 m in diameter and 50 m in height. The capacity of each silo is approximately 16,700 drums. The second stage of the LILW facility is under construction to be completed by December 2019. Site construction is currently underway. The second stage of the LILW facility is a type of near-surface disposal with disposal capacity of 32,875 m{sup 3} (125,000 drums). This stage of the LILW disposal facility consists of an underground gallery, three movable cranes and twenty vaults, etc. Vault for final disposal is located 70 m above sea level, 20 m in width, 20 m length and 10 m in height. The capacity of each vault is approximately 6,250 drums. These stages of the LILW disposal facilities are together called LILW complex disposal facility. It's the world's first LILW disposal facility with two different types of disposal facilities in one site. For efficient operation, these disposal facilities share a reception and storage building, waste water treatment system, measurement and monitoring system, etc. After construction of the first two stages is completed, LILW disposal facilities will be incrementally expanded, and as the whole LILW disposal facilities are completed, all LILW in Korea will be managed safely in WLDC. (authors)

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DOE OSTI2016

Evaluation and Impacts of Mercury in the SRS Liquid Waste System - 16121

The Savannah River Site (SRS) Liquid Waste System (LWS) stores radioactive waste in 43 underground tanks. The radionuclides in the waste are removed through a series of separation processes and the low-level fraction is immobilized in a grout waste form while the high level fraction is disposed in a glass waste form. Mercury originated from decades of canyon processing (used as a catalyst for dissolving the aluminum cladding of reactor fuel) and is present throughout the LWS (∼60 metric tons). Mercury has long been a consideration in the LWS, from both hazard and processing perspectives. Mercury is removed from the LWS in many ways including: 1) the Tank Farm evaporator condensates; 2) at the Defense Waste Processing Facility (DWPF) where it is steam-stripped and removed during the feed preparation processes; 3) it is removed at the Effluent Treatment Plant (ETP); and it is immobilized in the low-level, grout waste form. An integrated, system-wide evaluation and assessment of mercury behavior in the LWS is being performed. The recently completed Phase I evaluation focused on mercury inventory and speciation in the LWS; mercury holdup and chemical processing behavior; mercury impact Identification, including worker safety and equipment degradation; and mercury removal and disposal options. The mercury removal system at the DWPF has been unable to remove mercury successfully due to equipment and chemistry challenges since the 2008 timeframe. Phase I evaluations indicate that approximately 43 percent of the mercury in waste sent to the DWPF is now returning to the Tank Farm in the recycle stream and is concentrating to higher than expected levels (∼500 mg/l) around the 2H evaporator system. The chemical behavior of mercury is further complicated by the formation of higher than expected quantities of organomercury species, specifically methyl mercury, now present in the DWPF recycle evaporator system (2H Tank farm evaporator), which is used to volume-reduce the recycle stream. Phase II activities are building on the Phase I mercury activities, including additional sampling and characterization activities, a re-assessment of overall system knowledge, ranking and prioritizing critical gaps/information in mercury behavior across the flowsheet, assessing impacts of mercury removal and disposal options, and documenting an action plan for overall mercury management in the LWS. Phase II activities include two System Engineering Evaluations to 1) 'Reestablish Mercury removal Capability within DWPF', and 2) 'Determine the Best Alternative Mercury Removal Location with the LWS'. This paper provides the results of the Phase I and II mercury evaluations for the SRS LWS. (authors)

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DOE OSTI2015

Addressing an aging facility: improving reliability of the defense waste processing facility at the Savannah River Site - 15539

The High Level Waste program at the Savannah River Site significantly reduces environmental risks associated with the storage of radioactive waste from legacy efforts to separate fissionable nuclear material from irradiated targets and fuels. The Defense Waste Processing Facility treats and immobilizes High Level Waste into a durable borosilicate glass for safe, permanent storage. Construction of the facility began in 1983 and radioactive operations commenced in 1996. In an effort to support the disposition of radioactive waste and accelerate tank closure for the balance of the Savannah River Site mission, the Defense Waste Processing Facility recently underwent an extensive reliability review in 2013 to understand future needs as well as to address aging infrastructure. As a product of this review, a Performance Improvement Plan was developed to systematically address three major focus areas including, a) asset preservation, b) facility housekeeping, and 3) process improvements. Since issuance of the plan, significant progress has been made in each of these areas to improve the reliability and predictability of the Defense Waste Processing Facility. Providing consistent, constructive, and concise communications is vital in the face of a changing budgetary environment. The Savannah River Site case study provides an example of the U.S. Department of Energy - Savannah River (DOE-SR) and SRR partnering to inform stakeholders about cleanup progress at the site. DOE-SR and SRR periodically and consistently provide both annual written progress reports under the Federal Facility Agreement and quarterly written progress reports under the landfill permit to the government regulators. Through the Liquid Waste System Plans, DOE-SR and SRR constructively and concisely summarize both the significant progress made and explain how a changed budget can impact the goals moving forward. Furthermore, DOE-SR and SRR informs the community and general public about the Site's progress and future during Citizens Advisory Board meetings. In addition, through the new proposed model for complex-wide communications, stakeholders can better understand how budget constraints will affect them. The set-aside program would also encourage stakeholder participation and involvement in helping each state and site obtain extra funding through performance evaluations. Waste management organizations can benefit in a number of ways from an effective partnership approach to communicate with stakeholders about available funding. Keeping an open and honest dialogue with regulatory agencies can help preserve confidence, thus mitigating or completely avoiding costly and distracting fines and penalties. Informing the public not only promotes trust, but also spurs the public to exercise its right to demand more information and voice any concerns to elected or DOE officials. With this approach, DOE can receive better understanding and cooperation from stakeholders in a changing budgetary landscape. Furthermore, a set-aside competitive program would encourage efficiency and innovation that could lead to further cost savings for the government while providing extra funding for cleanup efforts. (authors)

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DOE OSTI2016

Using Green's Functions to Compute 3 Dimensional Neutron Flux Maps - 16016

The knowledge of the radionuclide content of radioactive waste is of utmost importance for safety and waste management reasons. Numerical simulations are used by EDF-DP2D to anticipate dismantling and radioactive waste management. The main numerical activation schemes include the computing of a 3 dimensional neutron flux map, obtained on the basis of a neutron propagation calculation. The codes used solve the transport equation called the Boltzmann equation without simplification. The Monte Carlo method is used to solve the Boltzmann equation. Random series of numbers are used to simulate the lives of millions of neutrons. The codes follow each neutron individually, from its birth to its disappearance by leakage, absorption, or fission, and then the neutron flux map is calculated at the nominal power rating conditions. The main drawback of the Monte Carlo codes is the large computing time needed to converge within acceptable statistical criteria. This extended time period is incompatible with the multiple fuel managements used by EDF to operate its NPPs because it is necessary to make a Monte Carlo calculation for each fuel loading pattern. The solution developed by EDF-DP2D is to use the importance factors of Green's Functions to simplify the computing of the 3 dimensional neutron flux maps. Physicists talk about Green's Functions where mathematicians talk about elementary solutions of linear equations with constant coefficients. With an activation scheme based on the Green's Functions approach the 3D neutron flux is obtained on the basis of the convolution of the normalized source contributions (i.e. normalized response matrices) with the neutrons emitted by the fuel assemblies. The response matrices to unitary sources are computed with the Monte Carlo code MCNP for different tallies of interest and the neutron sources emitted by the fuel are computed with a core code for different fuel loading patterns. The principal interest of this approach is the calculation of the response matrices to unitary sources that can be used for all power distributions, thus limiting the Monte Carlo computing time. (authors)

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DOE OSTI2015

A Preliminary Analysis on the Applicability of Deep Borehole Disposal for Spent Fuels in Korea - 15496

A deep geological disposal system, the disposal depth is about 500 m below ground, is considered as the safest method to isolate the spent fuels (SF) or high-level radioactive waste from the human environment with the best available technology at present time. The disposal safety of this system has been demonstrated with underground research laboratory and some advanced countries such as Finland and Sweden are implementing their disposal project on commercial stage. However, if these high-level radioactive wastes can be disposed of in deeper and more stable rock formation than deep geological disposal depth, it has several advantages. Therefore, as an alternative disposal concept, i.e., deep borehole disposal technology is under consideration in number of countries in terms of its outstanding safety and cost effectiveness. In this paper, the general concept for deep borehole disposal of spent fuels or high level radioactive wastes was reviewed. And the key technologies and challenges in development of this disposal method with the nuclear environment were analyzed. Also, very preliminary deep borehole disposal concept including disposal canister concept was developed according to the nuclear environment. These results will be used as an input for the analyses of applicability for DBD in Korea. (authors)

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DOE OSTI2019

Oxidation of Organics in the Defense Waste Processing Facility Condensate

Defense Waste Processing Facility (DWPF): The state of South Carolina currently holds 35 million gallons of radioactive waste in 51 underground tanks. DWPF was built in an effort to convert the waste into a more stable form. It turns the sludge waste from the tanks into glass, which will keep the waste stable for 10,000 years. Currently, DWPF uses formic acid in its flowsheet and there is one major problem associated with its use: it decomposes into hydrogen, a flammable gas. Glycolic acid, however, has been found as a possible alternative that performs similarly to formic acid but does not directly form hydrogen. One of the main issues with the glycolic acid flowsheet is the potential for there to be glycolate and other organics in DWPF's Recycle Collection Tank (RCT). These organics could lead to hydrogen generation later in the downstream evaporators and must be destroyed in the RCT in order for glycolic acid to be a suitable replacement for formic acid. Oxidants can be added to the RCT to destroy the organics. Two oxidizers, permanganate and Fenton's reagent (iron catalyzed hydrogen peroxide), are possible candidates. Find optimum processing conditions for the oxidation of organic species in a typical RCT simulant. A reactor served as a small-scale version of the RCT. During each run, an RCT simulant with 8 added organics, including glycolate, was created in the reactor. Either permanganate or peroxide was slowly added for each run through a syringe pump. Each experiment lasts 6 hours and a total of 23 runs are scheduled. Variables of Interest: Oxidizer (Permanganate or Peroxide); Temperature (15 deg. C to 50 deg. C); Oxidant addition time (10 mins to 2 hours); Amount of oxidant (1 x to 1.5 x stoichiometric requirements); Solution pH (3 to 13); Glycolate concentration (125 to 250 mg/kg). Offgas Analyzers: Mass spectrometer and FTIR spectrometer. Samples: pulled at 7 different times throughout the run to determine the amount of organic and inorganic carbon and concentrations of various organics through ion chromatography (IC). Only 3 of the runs have been completed at this point, all with permanganate. Based on the first 3 runs, permanganate seems to oxide the majority of the organics and the run at pH 8 and 32.5 deg. C had the most effective oxidation. The rest of the runs will need to be performed to determine the best oxidant and optimum values of the other variables of interest. Once the experiment is completed, DWPF will select one of the oxidants. Testing will continue to determine its performance during foam-over and other non-ideal conditions. Finally, the oxidant will be tested with actual radioactive waste to confirm its effectiveness.

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DOE OSTI2021

FOURTH QUARTER / ANNUAL TRANSPORTATION REPORT FISCAL YEAR 2021

Waste Shipments to and from the Nevada National Security Site, Radioactive Waste Management Complex. This report satisfies the U.S. Department of Energy (DOE) commitment to prepare a quarterly summary of waste shipments to the Nevada National Security Site (NNSS) Radioactive Waste Management Complex (RWMC) in Area 5.

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DOE OSTI2017

Thermal Testing of Waste Packages for a Storage Environment

The Pipe Overpack Container (POC) was developed at Rocky Flats to transport plutonium residues with higher levels of plutonium than standard TRU waste to the Waste Isolation Pilot Plant (WIPP) for disposal. The POC consists of an inner Pipe Container (PC) surrounded by organic fiber dunnage inside of a 7A drum. The PC was design to maintain separation of fissile material and to provide shielding from radiation. The POC has many design features that are similar to Type B packages that do provide protection from engulfing fires. However, it was not specifically designed to prevent release of stored material during a long-duration fire. POCs are now being used to store combustible TRU waste at Department of Energy (DOE) sites (e.g., Waste Isolation Pilot Plant). Radioactive waste storage operators would like to be able to claim some level of protection is provided by the POCs for thermal assaults that could occur within storage facilities when these POCs contain PCs with combustibles. In 1997 Sandia National Laboratories (SNL) conducted a series of tests to determine the degree of protection POCs provide during storage accident events. One of these tests exposed four of the POCs to a 30-minute engulfing pool fire, resulting in one of the 7A drums generating sufficient internal pressure to pop off its lid and expose the top of the PC to the fire environment. However, these tests are not bounding for POC/PC containing combustibles. With the addition of combustible materials in the PC, the possibility exists for generation of higher quantities of gases and much higher pressures inside the PC, which can lead to early rupture of the PC filter and higher radioactive aerosol release fractions. Therefore, a new series of tests was required to determine the temperature response of the exposed PC and determine the amount of material that could be released. DOE contacted SNL in 2015 to discuss the 1997 SNL tests and came to the conclusion that those tests were not bounding for POC/PC containing combustibles Therefore, DOE commission a new set of tests at SNL. Starting in 2015, SNL conducted fire tests to examine whether the contents of the PC would reach a temperature that would result in generation of sufficient gas to cause overpressurization of the PC, and subsequent release of its content. This paper will describe the various test conducted thus far, and present preliminary results from these tests.

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DOE OSTI2015

Decommissioning of the NPP A-1 heavy water management system - 15121

The paper deals with experience and techniques in the application of decontamination technology and remotely controlled robotic devices for the decontamination and dismantling of the heavy water management system during undergoing decommissioning process of the A-1 NPP as well as treatment of arising liquid waste All of these activities are characterized by high level of radioactivity and contamination. During operation of the A-1 NPP the heavy water management system fulfilled several functions: (1) cooling of the moderator, (2) topping up of the moderator in the active zone, (3) drainage, feed and storage of heavy water, (4) capture of leaked heavy water. Closely associated with heavy water management were systems for separation and isotopic purification of heavy water (distilling station, removing organic impurities station, isotopic purification station). The NPP heavy water system is located inside the power block building in several rooms where the inner surface contamination is up to the level of 10 E4 Bq.cm{sup -2}, dose rate up to 15,0 mGy.h{sup -1} and the feeding pipeline contained liquid RAW with high tritium content. Pre-dismantling decontamination of the heavy water management system and associated systems had to be performed with the intention of improving the radiation situation, i.e. reducing contamination of internal surface and dose rates in the vicinity of the equipment. Various sophisticated procedures such as retrieval of sludge using remotely operated devices and loop chemical decontamination of internal surfaces were applied. The generated radioactive waste such as spent decontamination solutions and sludge were immediately solidified on-site into a geo-polymer matrix using a mobile conditioning facility. For dismantling and fragmentation of some devices, mobile robotic systems MT80 and MT15 were applied which had been developed, designed and constructed as general-purpose decommissioning equipment. Special tooling was developed for application with the robots, such as hydraulic shears, circular saw, reciprocating saw, circular pipe cutter and a system for quick tool-change without direct intervention of the operators. Most of the operations are remotely controlled on basis of visual information from four cameras, with consistent radiation protection of the operators. The implementation of partial decommissioning tasks in such hard and unfavorable radiation conditions (moreover accompanied by generation of specific high-content alpha radwaste) has introduced in practice several new unique decontamination and dismantling techniques and new procedures for the effective on-site processing of generated secondary radioactive waste, including sludge and sorbents. These advantageous techniques, equipment and experience will be exploited in the continuation of the A-1 NPP decommissioning process which has been organized according to distinctively licensed phases and are also being applied in the decommissioning of other NPP in Slovakia as well as abroad. (authors)

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DOE OSTI2015

Overview of the DOE nuclear fuels storage and transportation planning project - 15444

The US Department of Energy (DOE) Office of Nuclear Energy is conducting planning activities within the Nuclear Fuels Storage and Transportation Planning Project (NFST) to lay the groundwork for implementing interim storage, including associated transportation, per the Administration's Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste, and to develop a foundation for a new nuclear waste management organization. The purpose of the NFST is to make progress on this important national issue within existing legislative and budgetary authorizations. The objective of the NFST is to develop and begin implementation of an integrated plan to (1) implement interim storage; (2) improve the overall integration of storage as a planned part of the waste management system; (3) prepare for the large-scale transportation of spent nuclear fuel and high-level radioactive waste, with an initial focus on removing SNF from the shutdown reactor sites; and (4) develop foundational information, resources, and capabilities needed to support the aforementioned objectives and future implementation decisions and actions. An over-arching goal is to develop options for decision-makers on the design of an integrated waste management system. The planning activities are divided into four major topical areas: (1) Consent-Based Siting, (2) Storage, (3) Transportation, and (4) Strategic Crosscuts. This paper provides an overview of the NFST and its activities. (authors)

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DOE OSTI2019

Panel Session 16: Hot Topics and Emerging Issues in US Commercial LLRW Management and Disposal

This panel focused on emerging issues in commercial LLW management in the US from the perspective of active members of the Low-Level Radioactive Waste Forum, Inc. State, compact, federal and industry officials will share their views on a variety of timely and significant topics related to low-level radioactive waste management, disposal and related issues. Presentations proposed include: 1 - Overview of NRC Activities and Initiatives (John Tappert, US NRC (USA)); 2 - Advocates for Responsible Disposal in Texas (Edward Selig, Advocates for Responsible Disposal in Texas (ARDT) (USA)); 3 - Hot Topics and Emerging Issues in U.S. Commercial LLW Management and Disposal (Joe Weismann, US Ecology (USA)); 4 - DOE/NNSA Office of Radiological Security (ORS) Sealed Source Management and Disposal (David Martin, Energetics Incorporated (USA), Temeka Taplin, US DOE NNSA (USA))

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DOE OSTI2015

Understanding the consequences of hypothetical post-closure transient criticality in a geological disposal facility - 14439

Radioactive Waste Management (RWM) is responsible for implementing geological disposal of the UK's higher-activity radioactive wastes. These wastes include significant quantities of fissile nuclides and, therefore, demonstration of criticality safety for the wastes under disposal conditions forms an important component of RWM's disposal system safety case. At the time of disposal all waste packages will be safely sub-critical. Containment provided by the engineered barrier system of a geological disposal facility (GDF) will ensure that criticality is not possible for some time after disposal. However, as waste packages degrade, fissile and other materials may be mobilised, which could affect the potential for criticality. Although the likelihood of a critical configuration developing is demonstrably low, it cannot be assumed to be zero over long post-closure timescales. The UK environment agencies' Guidance on Requirements for Authorisation for a GDF includes a specific requirement to consider the consequences of hypothetical 'what-if' post-closure criticality. RWM has undertaken a programme of research to model and understand the consequences of hypothetical post-closure criticality events. Initially focused on the scenario of fissile material relocation and accumulation following intermediate level waste (ILW) disposal, two bespoke computer models were developed. The quasi-steady-state (QSS) model considered critical systems maintained by accumulation of fissile materials where the temperature feedback remains negative throughout the transient. The rapid transient model (RTM) considered systems with positive temperature feedback. The QSS model development included a successful comparison with the Oklo natural reactors. With a requirement to consider post-closure criticality for a wider range of potential waste types under different illustrative disposal concepts, the initial transient criticality models had some limitations. For the QSS model this was the assumption of fixed geometry, while the complex functional inputs required in the RTM were not readily available for different materials. For the QSS model this paper describes the model development to allow changing volume; this has enabled its application to 'what-if' scenarios of flooding within a spent fuel waste package and the slumping of a stack of ILW packages. For rapid transient criticality this paper explains how the development of a simpler 'bounding approach' model to complement the RTM has enabled transient criticality analysis for a wider range of host materials. The role of this research in demonstrating post-closure criticality safety of a GDF in the UK is presented in a companion paper which also draws on the results of a parallel research project to demonstrate that the likelihood of post-closure criticality is low. (authors)

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DOE OSTI2016

Secondary Waste Form Development and Data Package for Hanford's Integrated Disposal Facility Performance Assessment - 16479

The Hanford Low-Activity Waste (LAW) Vitrification facility is planned to immobilize Hanford tank waste for on-site, near-surface disposal at Hanford's Integrated Disposal Facility (IDF). During this process, liquid and solid secondary wastes will be generated which are also slated for disposal at the IDF. In order to receive authorization to dispose of the waste in the IDF, a Performance Assessment (PA) must be completed in accordance with DOE Order 435.1, Radioactive Waste Management, and its accompanying Manual. The PA must include calculations of potential doses to a member of the public for a 1,000-year period to provide a reasonable expectation that performance objectives will not be exceeded and that the waste will be managed in a manner that is protective of worker and public health and safety, and the environment. Liquid and solid secondary wastes are planned to be immobilized within cementitious waste forms similar to those used elsewhere in the United States and around the world for permanent disposal of low-level radioactive waste (LLW). Data packages and supporting documents are being developed to provide the technical underpinning for the analyses of contaminant releases to the environment from these waste forms for the IDF PA. Traditionally, releases from cementitious waste forms have utilized a diffusion-limited release model using contaminant-specific diffusivities measured in standard leach tests as inputs. More mechanistic modeling techniques are under development to increase the rigor of the PA analyses. In addition, recent testing has shown diffusivities of key contaminants can be reduced by several orders-of-magnitude through optimized formulation of the waste form and use of additives to bind specific contaminants. (authors)

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DOE OSTI2016

Evaluation of Current Knowledge for Building the Dutch Salt Safety Case - 16024

This paper provides an overview of an evaluation for building the Safety Case of geological disposal of radioactive waste in rock salt in the Netherlands. The evaluation comprises the present knowledge about the safety and feasibility of geological disposal designs considered in the past, thereby classifying the information taking into account the methodology and structure of a Safety Case. The evaluation is performed as part of the Dutch research programme OPERA (OnderzoeksProgramma Eindberging Radioactief Afval - Dutch acronym for Research Programme into Geological Disposal of Radioactive Waste) by a consortium consisting of GRS from Germany, and TNO and NRG from the Netherlands. Basis of the evaluation is the available Dutch and international scientific-technical information about geological disposal in rock salt, especially from the German and US programmes. The evaluations have shown that although much knowledge on the topic exists, the available information need to be integrated in order to develop a Dutch Safety Case on rock salt. For example, the characteristics of the Dutch radioactive waste investigated for disposal have changed considerably since the last Dutch research programme ended in 2000. This implies that the previously considered Dutch facility concepts for geological disposal in rock salt need to be adapted to the newly regarded waste types such as spent fuel from research reactors and large amounts of LLW, including depleted uranium. Introducing these aspects into the iteration towards a salt Safety Case would also require an update of the post-closure safety assessment. (authors)

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DOE OSTI2018

Andra's Underground Research Laboratory as a Tool to Prepare Monitoring System of Cigeo Project - 18070

Andra, as French radioactive waste management agency, plans to build an underground disposal facility, named Cigeo, dedicated to high and intermediate level, long lived radioactive wastes. To ensure the safety during the disposal operations that includes the principle of reversibility, the design of the associated monitoring system will be based on specific research and development program and feedback obtained over more than 15 years on tests performed in representative conditions. Andra's Underground Research Laboratory (URL), located at 500 m deep, is excavated in a clay-rich formation. It is dedicated to (i) assess the thermo-hydro-mechanical, chemical and radiologic behaviors of the rock and supporting layers (concrete and steel), (ii) test envisaged monitoring system, (iii) provide feedback on operational organization and safety management and (iv) qualify operating methods (digging, concrete pouring, monitoring system design, etc.) to be implemented in the future disposal. Data collected from (i) characteristics of samples retrieved from boreholes, (ii) experts' observations, (iii) sensors installed in situ and (iv) manual measurements, are integrated in a database. This database is dedicated to samples management (more than 73 600 classified samples), data measurements collection, and conversion to physical values and available to plots or exports. Each day, approximately 1.6 million of values are integrated in the database. Finally, a geo-referenced database is linked to the main database in order to allow the spatial localization, both on the surface or in the underground galleries. All these tools give confidence in Andra's capacity to monitor Cigeo and be ready for the future industrial pilot phase. The monitoring system will also allow, step by step, optimizing the disposal design, performing regular safety control, helping in decision making for reversibility and improving long term safety. (authors)

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DOE OSTI2015

Analysis of the likelihood of post-closure criticality in a geological disposal facility - 14349

Radioactive Waste Management (RWM) is responsible for implementing geological disposal of the UK's higher activity radioactive wastes. These wastes include significant quantities of fissile nuclides and, therefore, demonstration of the criticality safety of the wastes under disposal conditions forms an important component of RWM's disposal system safety case. At the time of disposal, controls will ensure that all waste packages are sub-critical, and the containment provided by the engineered barrier system of a geological disposal facility (GDF) will ensure that criticality is prevented for some time after disposal. However, as waste packages begin to degrade, fissile and other materials may be mobilised, which could affect the potential for criticality. Therefore, the possibility for the evolution of conditions in a GDF to result in criticality requires assessment. RWM established the 'Likelihood of Criticality' research project to develop qualitative and quantitative arguments to demonstrate that criticality after closure of a GDF is unlikely. The project included consideration of the criticality safety of disposal concepts for intermediate and high level waste, spent fuel, separated plutonium and highly-enriched uranium in different generic geological settings. The results are underpinned by the construction of post-closure criticality scenarios that have been evaluated using high-level judgments about scenario credibility as well as modelling of evolving conditions in a GDF. In the latter case, a probabilistic modelling approach has been undertaken to address uncertainty in the calculation of concentrations and masses of fissile material that could occur at potential accumulation sites. The potential for criticality has been judged based on comparisons of the calculated fissile masses with minimum values required for criticality at different locations in the barrier system. This paper explains the novel approach taken to evaluating the likelihood of criticality in a GDF based on the probabilistic modelling of waste package evolution and plutonium and uranium migration. The role of this research in RWM's work to demonstrate the criticality safety of a GDF in the UK is presented in a companion paper, which also draws on the results of a parallel research project that demonstrates that, even if a criticality were to occur, then its consequences would be insignificant to GDF performance. (authors)

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DOE OSTI2018

Pathways towards Valid Data: Assessment of Optimized Packaging Strategies for Core Components - 18158

Every company that runs nuclear power plants needs a strategy for disposal of the radioactive waste and for decommissioning and dismantling of its nuclear reactors. The dismantling progress of several German NPPs is already in an advanced stage. Hence, particularly highly activated core components have to be disposed of as radioactive waste. Nevertheless, this is also an issue during the operating phase due to necessary replacements of highly activated core components such as PWR control elements or RPV heads. The conditioning of waste packages for final disposal in the German repository Konrad requires the core components to be dismantled, packaged and dried. Therefore, a qualified packaging strategy is needed to ensure a reliable compliance with the protection objectives of the repository and an optimized utilization of prescribed contingents. Exemplified by PWR control elements a pathway is drawn how to perform an optimized packaging strategy in compliance with the protection objectives of the repository based on only few relevant radionuclides derived from activation calculations. For these calculations, particularly the influence of trace impurities in the materials has to be considered and their amounts have to be verified or analyzed by sampling. Furthermore, low nuclide activities from contamination cannot simply be neglected because some of them are obliged to be declared or their radiological relevance has not been sufficiently determined. Due to various influence factors, qualified sampling strategies need carefully considered concepts. Therefore, detailed planning facilitates the qualification and approval process and reduces the costs for analyses and the number of containers. (authors)

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