Reactor Site Research Articles

As-Loaded Criticality Margin Assessment of Dual-Purpose Canisters Using UNF-ST&DARDS

A novel assessment has been completed to determine the previously unquantified and uncredited criticality margin available in as-loaded commercial spent nuclear fuel (SNF) canisters. This assessment was performed as part of a broader effort to assess issues and uncertainties with storage, subsequent transportation, and final disposal of SNF canister systems. Detailed analyses crediting the burnup, initial enrichment, and postirradiation cooling time of actual SNF inventory were performed for 554 SNF canisters stored at 23 commercial reactor sites to determine realistic criticality safety margins. These detailed analyses were automated by the Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS), a comprehensive, integrated data and analysis tool. Calculated, uncredited criticality margins (Δkeff) with respect to the safety analysis results range from 0 to almost 0.30 Δkeff for normal storage and transportation cases. Calculated eigenvalues (keff) range from 0.72 to 1.11 assuming a degraded neutron absorber disposal condition, and they range from 0.94 to 1.20 assuming a degraded basket disposal condition. Calculations with NaCl present in the moderator (which is possible for certain disposal geologies) were used to demonstrate the possibility for subcriticality for degraded cases with a keff above 0.98 with freshwater. The methods used to calculate keff for the canisters analyzed in this work are discussed in detail.The results demonstrate that, for the majority of canisters analyzed here, significant uncredited safety margin is available that could be used to compensate for uncertainties in the SNF assembly and canister internal components. These uncertainties are associated with long-term storage and subsequent transportation and disposal. Results also suggest that the inherent margins associated with how canisters are loaded could support future changes in licensing SNF storage and transportation systems to directly or indirectly credit the margins associated with actual SNF characteristics. Ongoing research continues to gather additional data to quantify uncredited safety margins for SNF canisters loaded at other nuclear reactor sites and to explore potential methods for applying this uncredited margin.

Transfer of natural radionuclides from soil to plants in Savar Dhaka

The radioactivity of environmental samples from nuclear reactor sites must be analyzed before the public is given free access to the plants grown in these soils. Plant and corresponding soil samples were collected from a sample site around the Savar research reactor near Dhaka (Bangladesh) and the activity concentrations of natural radionuclides <sup>226</sup>Ra (<sup>238</sup>U-chain), <sup>228</sup>Ra (<sup>232</sup>Th-chain) and non-chained <sup>40</sup>K were measured using gamma ray spectrometry. Soils of Savar contained more radioactive <sup>40</sup>K than <sup>226</sup>Ra and <sup>228</sup>Ra. The influence of certain soil properties on the activity concentrations and transfer factors (TF) of natural radionuclides were investigated by correlating the observed data with those of soil properties. The activity concentrations of <sup>40</sup>K were much higher than those of <sup>226</sup>Ra and <sup>228</sup>Ra in plants due to higher uptake from soils. The transfer factors for <sup>226</sup>Ra, <sup>228</sup>Ra and <sup>40</sup>K were found to range from 0.04 to 0.10, 0.12 to 0.32, and 0.24 to 0.72, respectively. The soil to plant transfer factors for <sup>40</sup>K was found to be much higher in plants, which might be due to this element being vital in plants. This study showed that activity concentrations of these radionuclides in plants and their plant transfer factors seem to depend on the activity concentrations of the same radionuclides in soil.</p><p><strong> </strong>

Subsurface Temperature Modeling using Integrated Modeling for Nuclear Reactor Site Assessment in Volcanic Zone

Indonesia has giant vulcanic arc that almost the largest vulcanic arc in the world. Therefore, one of the main risk for nuclear site plan is the vulcanic area. Therefore to reduce the risk, one of most safety nuclear site plant is old vulcanic area. In this paper, we propose to predict subsurface temperature profile to ensure the condition of subsurface of vulcanic zone. Geothermal heat flow is important parameter in modeling of subsurface temperature. The subsurface temperature is one of vulcanic activity parameter which very important for nuclear site plant risk assesment. The integrated modeling for predicting subsurface temperature profile is carried out by combining geothermal heat flow and subsurface profiles resulted from either seismic or gravity measurement. The finite difference of Fourier’s law is applied to surface temperature, temperature gradient, geothermal heat flow and thermal conductivity profile for producing subsurface temperature distribution accurately. This subsurface temperature profile is essential to characterize the vulcanic zone whether it is still active or inactive. Characterization of vulcanic activity is very useful to ensure or to minimize the risk of nuclear site plant in vulcanic zone. One of interesting case study of nuclear site plan in Indonesia is mount Muriah site plan, this method is useful to ensure whether mount Muriah is still active or inactive now

Nuclear Waste Management Decision-Making Support with MCDA

The paper proposes a multicriteria decision analysis (MCDA) framework for a comparative evaluation of nuclear waste management strategies taking into account different local perspectives (expert and stakeholder opinions). Of note, a novel approach is taken using a multiple-criteria formulation that is methodologically adapted to tackle various conflicting criteria and a large number of expert/stakeholder groups involved in the decision-making process. The purpose is to develop a framework and to show its application to qualitative comparison and ranking of options in a hypothetical case of three waste management alternatives: interim storage at and/or away from the reactor site for the next 100 years, interim decay storage followed in midterm by disposal in a national repository, and disposal in a multinational repository. Additionally, major aspects of a decision-making aid are identified and discussed in separate paper sections dedicated to application context, decision supporting process, in particular problem structuring, objective hierarchy, performance evaluation modeling, sensitivity/robustness analyses, and interpretation of results (practical impact). The aim of the paper is to demonstrate the application of the MCDA framework developed to a generic hypothetical case and indicate how MCDA could support a decision on nuclear waste management policies in a “small” newcomer country embarking on nuclear technology in the future.

对我国AP1000核电离堆废物处理技术的对比分析

浙江三门核电厂和山东海阳核电厂采用美国开发研究的AP1000堆型第三代核电机组，首次采用了核岛废物处理系统与厂址废物处理设施(SRTF)相结合的处理模式，但两厂址废物处理设施(SRTF)处理工艺上存在一定的差异，本文针对两厂址废物处理设施(SRTF)处理工艺进行了对比分析。综合研究AP1000离堆废物处理技术，形成规范化的、系列化的处理技术及装备，对实现AP1000核电厂离堆废物的安全管理、落实废物最小化原则发挥积极作用。 Zhejiang Sanmen nuclear power plant and Shandong Haiyang nuclear power plant adopt the third-generation nuclear reactor named AP1000 which is developed by the United States. They are the first companies to use the combined process which combine with the nuclear waste disposal system and a site radwaste treatment facility (SRTF). However, the two plants have some differences in their SRTF processes. This paper contrasts and analyzes the processes of the two SRTF. It has played a positive role for making a comprehensive research on the treatment technologies of AP1000 reactor site radwaste and forming of standard and systematic treatment technologies and equipment. And it has a positive effect on realizing the safety management of the AP1000 reactor site radwaste and implementing the principle of waste minimization.

Small Modular Reactor Site Considerations for Uganda

Small Modular Reactor Site Considerations for Uganda

Reaction: Recycling and Generation IV Systems

Dame Sue Ion is chair of the UK Nuclear Innovation Research Advisory Board. She represents the UK on a number of international review and oversight committees for the nuclear sector, including the Euratom Science and Technology Committee, which she chairs. She is the only non-US member of the US Department of Energy's Nuclear Energy Advisory Committee, on which she has served since 2005.

The New Facilities for Neutron Radiography at the LVR-15 Reactor

Neutron radiography is an imaging method often used at research reactor sites. Back in 2011 a project was started with the goal to build a neutron radiography facility at the site of the LVR-15 research reactor in Rez, Czech Republic. In the scope of the project two horizontal channels were adapted for the needs of neutron radiography. This comprises the HC1 channel which offers an intense thermal neutron beam with a diameter of 10 cm, which can be used for imaging of larger samples, and the HC3 channel which beam is restricted just to 4x80 mm2, but is highly thermalized, collimated and reduced from gamma background, thus capable of providing better radiograph resolution. Both facilities are equipped with newest Timepix based detectors, with thin 6LiF converters for neutron detection capable of delivering high resolution. Both facilities offer a unique opportunity for non-destructive testing in the Czech region. In 2015 both facilities were put into test operation and several radiographs were acquired, which are presented in the following text.

Estimation of Inherent Safety Margins in Loaded Commercial Spent Nuclear Fuel Casks

A novel assessment has been completed to determine the unquantified and uncredited safety margins (i.e., the difference between the licensing-basis and as-loaded calculations) available in as-loaded spent nuclear fuel (SNF) casks. This assessment was performed as part of a broader effort to assess issues and uncertainties related to the continued safety of casks during extended storage and transportability following extended storage periods. Detailed analyses crediting the actual as-loaded cask inventory were performed for each of the casks at three decommissioned pressurized water reactor sites to determine their characteristics relative to regulatory safety criteria for criticality, thermal, and shielding performance. These detailed analyses were performed in an automated fashion by employing a comprehensive and integrated data and analysis tool—Used Nuclear Fuel-Storage, Transportation and Disposal Analysis Resource and Data System (UNF-ST&DARDS). Calculated uncredited criticality margins from 0.07 to almost 0.30 Δkeff were observed, calculated decay heat margins ranged from 4 to almost 22 kW (as of 2014), and significant uncredited transportation dose rate margins were also observed. The results demonstrate that at least for the casks analyzed here, significant uncredited safety margins are available that could potentially be used to compensate for SNF assembly and canister structural performance related uncertainties associated with long-term storage and subsequent transportation. The results also suggest that these inherent margins associated with how casks are loaded could support future changes in cask licensing to directly or indirectly credit the margins. Work continues to quantify the uncredited safety margins in the SNF casks loaded at other nuclear reactor sites.

"Janus" Calixarenes: Double-Sided Molecular Linkers for Facile, Multianchor Point, Multifunctional, Surface Modification.

We herein report the synthesis of novel “Janus” calix[4]arenes bearing four “molecular tethering” functional groups on either the upper or lower rims of the calixarene. These enable facile multipoint covalent attachment to electrode surfaces with monolayer coverage. The other rim of the calixarenes bear either four azide or four ethynyl functional groups, which are easily modified by the copper(I)-catalyzed azide–alkyne cycloaddition reaction (CuAAC), either pre- or postsurface modification, enabling these conical, nanocavity reactor sites to be decorated with a wide range of substrates to impart desired chemical properties. Redox active species decorating the peripheral rim are shown to be electrically connected by the calixarene to the electrode surface in either “up” or “down” orientations of the calixarene.

Results of the engineering run of the Coherent Neutrino Nucleus Interaction Experiment (CONNIE)

The CONNIE detector prototype is operating at a distance of 30 m from the core of a 3.8 GWth nuclear reactor with the goal of establishing Charge-Coupled Devices (CCD) as a new technology for the detection of coherent elastic neutrino-nucleus scattering. We report on the results of the engineering run with an active mass of 4 g of silicon. The CCD array is described, and the performance observed during the first year is discussed. A compact passive shield was deployed around the detector, producing an order of magnitude reduction in the background rate. The remaining background observed during the run was stable, and dominated by internal contamination in the detector packaging materials. The in-situ calibration of the detector using X-ray lines from fluorescence demonstrates good stability of the readout system. The event rates with the reactor ON and OFF are compared, and no excess is observed coming from nuclear fission at the power plant. The upper limit for the neutrino event rate is set two orders of magnitude above the expectations for the standard model. The results demonstrate the cryogenic CCD-based detector can be remotely operated at the reactor site with stable noise below 2 e− RMS and stable background rates. The success of the engineering test provides a clear path for the upgraded 100 g detector to be deployed during 2016.

The strange history of polarised neutrons in Australia

The history of polarised neutrons in Australia is unusual firstly because of the particular access that individuals in universities had to the facilities at the reactor site and because this resulted in the experiments being done almost all being with polarisation analysis. Two instruments were initially available. One was a conventional instrument albeit with a tilting counter. The other was a primitive polarisation analysis instrument purpose built for diffuse scattering. This latter instrument evolved over more than thirty years and produced results ranging from the separation of magnetic and nuclear diffuse scattering, for which it was conceived, to the isolation of magnetic features in inelastic spectra.

Cost Estimation and Efficiency Analysis of Korean CANDU Spent Fuel Disposal Alternatives in Consideration of Future Price Volatility

In Korea, spent fuel is temporarily stored in spent fuel pools at nuclear reactor sites and it is predicted to become saturated between 2020 and 2024. For this reason, four disposal alternatives (KRS-1, A-KRS-1, A-KRS-21, and A-KRS-22) have been developed in order to carry out the direct disposal of the CANDU spent fuel. The objective of this study is to conduct cost efficiency analysis of the disposal alternatives in consideration of price volatility for the radioactive waste repository. To derive future price volatility, this study used the ARIMA model. As a result, A-KRS-1 is the most efficient in terms of price per bundle using 2015 price. As for the results using ARIMA model, except in the case of KRS-1, the cost per bundle of A-KRS-1, A-KRS-21, and A-KRS-22 is decreased. Cost estimation using ARIMA model shows little change or decreases in cost while cost estimation using inflation rates for 2020 resulted in approximately 7.2% increases compared to 2015 for all options. As for the results of scenario analysis, A-KRS-1 earned 8,160 points, while A-KRS-22 followed closely behind with 7,980 points among the total 24,300 points. The results of this study provide invaluable policy data for any nation considering the construction of spent nuclear fuel repository.

‘You are the first journalist and you are the last journalist who will ever come here’: Nuclear secrets and media practices of access-trespass

In India, information about nuclear technologies is often kept shrouded in secrecy. Science reporters covering such strategic sectors depend for information on cultivating sources and pursuing contacts in the nuclear establishment. Based on in-depth interviews with two science/environmental journalists and analysis of their television reports and magazine articles, I show how journalists are acutely aware of their role as mediators between scientists and their complex technological projects on the one hand and the general public on the other. Reporting on ‘secret’ nuclear sciences makes concerns of objectivity and bias in journalistic practice strategic: if journalists are considered pro-nuclear, they have a better chance of accessing nuclear reactor and space research sites. Journalists and scientists co-design and co-stage experiments to be witnessed by the television audiences, and I argue for a close analysis of these mutual entanglements of scientific processes and media practices to understand the performative mediations of environmental debates. Furthermore, I examine how television studio and split-screen management affords news anchors a strategic advantage in confronting politicians and science experts with questions about the risk and safety of scientific projects – an advantage that is not equally available to journalists while accessing strategic technoscience sites.

Between Grassroots Protest and Green Politics: The Democratic Potential of the 1970s Antinuclear Activisim

This article narrates the development of the antinuclear movement from the bottom up, showing how local protests initiated changes in Germans' ideas about democracy and public participation, precipitating the Green Party's emergence. The narrative begins with the pre-history of the 1975 occupation of the Wyhl reactor site in Southern Baden. It shows that vintners' concerns about the future of their livelihoods underpinned protests at Wyhl, but argues that the anti-reactor coalition grew in breadth after government officials' perceived misconduct caused local people to connect their agricultural concerns with democracy matters. It then explains how local protests like the Wyhl occupation influenced the formation of the German Green Party in the late 1970s, showing how the sorts of convergences that occurred amidst “single issue” protests like the anti-Wyhl struggle enabled a wide variety of activists to come together in the new party. Thus, the article argues that particular, local concerns initiated a rethinking of participation in electoral politics. Far from fracturing society, these local concerns promoted diverse new coalitions and shaped an inclusive approach to electoral politics.

Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor

Fuel assemblies in the spent fuel pool are stored by suspending them in two vertically stacked layers at the Atucha Unit 1 nuclear power plant (Atucha-I). This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 wt% 235U and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. A gross defect detection tool, the spent fuel neutron counter (SFNC), has been used at the site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups, the existing signal processing software of the tool was found to fail due to nonlinearity of the source term with burnup. A new software package based on the LabVIEW platform was developed to predict expected neutron signals covering all ranges of burnups and cooling times. The algorithm employed in the software uses a set of transfer functions that are coupled with source terms based on various cooling times and burnups for each of the two enrichment levels. The software was benchmarked against an extensive set of measured data. Overall, out of 326 data points examined, the software data deviated from the measured data <10% in 87% of the cases. A further 10.5% matched the measurements between 10% and 20%. Thus, 97.5% of the predictions matched the measurements within the set 20% tolerance limit providing proof of the robustness of the software. This software package linked to SFNC will enhance the capability of gross defect verification at both levels in the spent fuel pool for the whole range of burnup, cooling time, and initial enrichments of the spent fuel being discharged into the various pools at the Atucha-I reactor site.

U.S. housing prices and the Fukushima nuclear accident

Did the nuclear catastrophe at Fukushima in March 2011 cause individuals to reappraise the risks they attach to nuclear power plants? We investigate the change in home prices in the U.S. after the Fukushima event to test the hypothesis that home prices in the proximity of power plants fell due to an updated nuclear risk perception. Using a difference-in-differences approach, we do not find evidence in support of the hypothesis that individuals reappraised the risks associated with nuclear power plants. According to our results home prices close to nuclear reactor sites did not fall relative to home prices at other locations in the U.S.

Dose assessment of radionuclides dispersion from Bushehr nuclear power plant stack under normal operation and accident conditions

Assessment of individual and collective doses around the unit-1 of Bushehr nuclear power plant (BNPP-1) site during normal nuclear reactor operation and under reactor accident conditions which are essential for safety and environmental analyses is presented in this paper. Release of radioactive materials to the environment following a design basic accident (DBA) (e.g. LBLOCA) is evaluated using HOTSPOT health physics computer code. In the normal operation conditions, atmospheric dispersion of radioactive material is calculated using CAP88-PC code. These codes utilize a Gaussian dispersion air transport plume model to simulate the atmospheric dispersion of radionuclides in different atmospheric stability classes and various wind speeds and directions. To calculate the doses under normal operation using CAP88 code, the surrounding area of the BNPP-1 within a radius of 30 km is gridded into 12 concentric rings and 16 sectors and the distribution of population and data of agricultural products are calculated for each grid. The meteorological data on atmospheric stability conditions, wind speed and the frequency distribution of wind direction based on data collected near the reactor site are also analyzed and applied. The results illustrate that the maximum total effective dose equivalent (TEDE) and committed effective dose equivalent (CEDE) values for personnel and public around the BNPP-1 site are lower than the annual effective dose limits for workers and public in normal operation and under accident conditions. Also, the results of the radiological impacts of the BNPP-1 in normal and accidental radioactive materials releases are compared with the dose values given in the Final Safety Analysis Report (FSAR) of the BNPP-1.

Transportation operations model analysis for removal of stranded fuel from shutdown reactors

The transportation operations model was used to identify options for removing stranded fuel currently in dry storage at nine shutdown reactor sites to a hypothetical consolidated storage facility. The logistical variables included the campaign duration, fuel selection priority, consist size and location of the consolidated storage and maintenance facilities. The major factors affecting the logistics of fuel removal were identified. Recommendations for optimal strategies for the transport of stranded fuel from shutdown sites are made.

SHIELDING ANALYSIS OF DUAL PURPOSE CASKS FOR SPENT NUCLEAR FUEL UNDER NORMAL STORAGE CONDITIONS

Korea expects a shortage in storage capacity for spent fuels at reactor sites. Therefore, a need for more metal and/or concrete casks for storage systems is anticipated for either the reactor site or away from the reactor for interim storage. For the purpose of interim storage and transportation, a dual purpose metal cask that can load 21 spent fuel assemblies is being developed by Korea Radioactive Waste Management Corporation (KRMC) in Korea. At first the gamma and neutron flux for the design basis fuel were determined assuming in-core environment (the temperature, pressure, etc. of the moderator, boron, cladding, UO2 pellets) in which the design basis fuel is loaded, as input data. The evaluation simulated burnup up to 45,000 MWD/MTU and decay during ten years of cooling using the SAS2H/OGIGEN-S module of the SCALE5.1 system. The results from the source term evaluation were used as input data for the final shielding evaluation utilizing the MCNP Code, which yielded the effective dose rate. The design of the cask is based on the safety requirements for normal storage conditions under 10 CFR Part 72. A radiation shielding analysis of the metal storage cask optimized for loading 21 design basis fuels was performed for two cases; one for a single cask and the other for a 2×10 cask array. For the single cask, dose rates at the external surface of the metal cask, 1m and 2m away from the cask surface, were evaluated. For the 2×10 cask array, dose rates at the center point of the array and at the center of the casks’ height were evaluated. The results of the shielding analysis for the single cask show that dose rates were considerably higher at the lower side (from the bottom of the cask to the bottom of the neutron shielding) of the cask, at over 2mSv/hr at the external surface of the cask. However, this is not considered to be a significant issue since additional shielding will be installed at the storage facility. The shielding analysis results for the 2×10 cask array showed exponential decrease with distance off the sources. The controlled area boundary was calculated to be approximately 280m from the array, with a dose rate of 25mrem/yr. Actual dose rates within the controlled area boundary will be lower than 25mrem/yr, due to the decay of radioactivity of spent fuel in storage.