bluuebunny/crossref_metadata_2025_split huggingface dataset:
Transient Modeling of Advanced High Temperature Reactor (AHTR) in RELAP5/SCDAPSIM/MOD 4.0
Authors: Lin Hsun-Chia , Zhang Sheng , Shi Shanbin , Sun Xiaodong , Christensen Richard 
Published: October 2018

The Advanced High Temperature Reactor (AHTR) is a fluoride-salt-cooled high-temperature reactor (FHR) design concept that is currently being developed at Oak Ridge National Laboratory for efficient production of electricity with improved safety features. Transient analyses of different scenarios are critical to demonstrate the safety of the AHTR design. An AHTR reactor model has been developed using RELAP5/SCDAPSIM/MOD 4.0. Thermodynamic and transport properties of three molten fluoride salts, namely FLiBe, FLiNaK, and KF-ZrF4, have been implemented into the RELAP5 code. The AHTR RELAP5 model consists of a reactor core, an upper plenum, a lower plenum, three primary loops, and three Direct Reactor Auxiliary Cooling Systems (DRACS) loops. DRACS Heat Exchangers (DHX) and Natural Draft Heat Exchangers (NDHX) are important components of DRACS and provide coupling between the primary loops and DRACS loops, and DRACS loops and air chimneys, respectively. Single-wall fluted tube heat exchanger designs have been proposed for the DHX and the NDHX to improve heat transfer performance in the two heat exchangers, and heat transfer correlations for fluted tubes have also been implemented into the RELAP5 code.
In this study, steady-state reactor normal operation and two transient scenarios are analyzed with the RELAP5 AHTR model. Based on a thermal hydraulics Phenomena Identification Ranking Table (PIRT) exercise, loss of forced circulation (LOFC) and loss of multiple DRACS loops are selected as the two transients for analysis. During transients, the decay heat is removed by the ambient air, fully relying on natural circulation/convection. The results of both transient scenarios show sufficient decay heat removal capabilities of DRACS with the proposed design.
Numerical and Experimental Investigation on Core Assembly Thermal-Gradient-Induced Deformation of Sodium-Cooled Fast Reactor
Authors: Ma Zehua , Wu Yingwei , Su G. H. , Tian Wenxi , Qiu Suizheng 
Published: October 2018

In sodium-cooled fast reactor (SFR), thermal gradient is the paramount factor of assembly transient bowing, that may cause great reactivity change, accelerate wrapper vibration wear, hindering the motion of control/shutdown rods, or worse yet, threatening the integrity of assemblies. However, because of the complexity of multi-assembly contact and interaction problem, it is difficult to assess the impact of core deformation on reactor performance safety.
The Core Assembly Deformation Test Facility (CADTF) is designed to perform a series of thermal bowing tests by Xi‘an Jiao Tong University (XJTU) to investigate the core deformation behaviors under thermal gradient. In this paper, a finite element model was established to simulate the mechanical response of single assembly under different flat-to-flat thermal gradient. The single assembly restrained bowing test performed in CADTF is chosen to validate the model. In the model, the measured temperature distribution as well as temperature-dependent elastoplastic and thermal expansion properties were taken into consideration. To ensure the model reliability, iterative computation is conducted by adjusting the friction coefficient of the load pads to match the calculated and measured contact force. According to the results, it can be seen that the three-dimensional displacement of assembly shows relatively good agreement with the experimental data. Therefore, it can be concluded that the model is capable of performing core deformation analysis for SFR.
The Role of the NRC in License Renewals
Authors: Miranda Samuel 
Published: October 2018

So far, the licenses of more than 85% of US operating plants have been renewed, authorizing them to continue operations for an additional 20 years past the end of their original, 40-year operating licenses.
10 CFR §54, which governs the NRC’s renewal of plant operating licenses, defines the Current Licensing Basis (CLB), and its role in license renewal applications. The CLB includes, inter alia, compliance with all the regulations in 10 CFR §50, and its appendices, including the design-basis information presented in final safety analysis reports (FSARs). Appendix A of 10 CFR §50 specifies the General Design Criteria (GDCs), and defines the conditions under which they must be satisfied. For example, Appendix A defines anticipated operational occurrences (AOOs) as incidents that may occur during the lifetime of a particular plant. Two standards of the American Nuclear Society (ANS) [1] [2] redefined AOOs as incidents that may occur during the calendar year for a particular plant, and added a class of events that may occur during the lifetime of a particular plant.
The ANS standards defined a categorization scheme that puts all the various types of incidents into four categories (i.e., Conditions: I, II, III, and IV), and specifies the GDCs and other criteria that must be met for each of the categories. AOOs, or Condition II events, for example, must not result in any fuel damage. Licensees have committed to abide by the ANS categorization system, and to comply with all of the categories’ acceptance criteria. These commitments are in their Final Safety Analysis Reports (FSARs), which are part of their CLBs.
Conditions I, and II define relatively frequently occurring incidents, and require that their consequences to be benign. Condition III events, however, are limited to only a very few incidents during the lifetime of a plant. Their consequences are not benign. (Condition IV is the most severe category. Condition IV events are not expect to occur at all.)
Since the frequency of occurrence of Condition III events is defined in terms of plant lifetime, it follows that lengthening the plant lifetime (e.g., from 40 to 60 years) will lead to the occurrence of more Condition III events. These events can result in fuel damage, or worse. The CLB of a renewed license plant, therefore, will have to account for more Condition III events.
This paper focuses upon how Condition III events can affect the CLB during an extended plant lifetime. It also discusses the concept of extending plant operating licenses by 20, or 40 years, and its potential impact upon the public health and safety.
Modification of Filtered Air Intake Flowrate to Improve Control Room Habitability
Authors: Liu Xinjian , Shu Weipeng , Wang Mengxi 
Published: October 2018

Control room habitability (CRH) shall be maintained to provide adequate protection for control room operators, such that they can remain in the control room envelope (CRE) safely for an extended period and thus control the nuclear facility during normal and accident conditions. A critical objective of CRH systems is to limit operator doses and/or exposure to toxic gases. The CRH systems does this by the combination of the intake of filtered air, isolation of outside air, recirculation systems and etc.
Among the parameters determining radioactivity in a control room (in proportion to radiation doses of operators), intake flowrate of filtered air is an important one. For different types of accident source terms, the evolution of operator doses in a control room versus intake flowrate were analyzed in this paper. It turns out that the increase of intake flowrate results in larger operator doses when inert radioactive gases are the dominant radioactive substances. On the contrary, increasing intake flowrate does good to lower the irradiation level of control room operators when radioactive aerosols dominate the source terms. The rationality behind this fact was interpreted in detail in this paper, with special attention paid to the unfiltered in-leakage rate. It can be inferred that an optimal intake flowrate probably exists leading to the minimum operator dose under an actual accident condition.
This paper then performed a calculation analysis based on design parameters and source terms of design basis accident of LOCA (a large break loss of coolant accident) accident. The evolution of operator dose was found to be a U-curve versus increasing intake flowrate, which proved the existence of the abovementioned optimal intake flowrate of filtered air for CRH systems. Furthermore, the sensitivity analysis of intake flowrate was carried out to study the effects of unfiltered in-leakage rate and filtered recirculation.
This study indicates that intake flowrate of filtered air can significantly influence the CRH. For different accidents, the intake flowrate should be properly modified rather than set as a fixed value. To optimize the radiological habitability of control rooms, the effects of unfiltered in-leakage must be taken into consideration. Besides, filtered recirculation is an effective way to control radiation exposure caused by iodine and radioactive aerosols.
A New Method of Integrating the RELAP5 to the RINSIM Simulation Platform
Authors: Tan Chao , Quiroga Victor M. , Fu Zheng , Xie Zhengquan 
Published: October 2018

This paper introduces a method to control the calculation progress of the RELAP5 codes and integrate them with other codes by interacting boundary data (such as general tables, TDVs and TDJs) at each step.
This work is basically finished with the support of the RINSIM simulation platform. The paper gives a brief introduction on RINSIM that how it controls the codes progress, sends the codes control commands, shares the values of different codes’ common blocks or modules.
However, the work can’t be done by just using the RINSIM, it also needs to modify RELAP5 codes. With the codes’ modification of commands responding, data reading/writing interface, data interacting interface, time step control and so on, we can build interface subroutines to integrate codes onto the RINSIM.
At the end, the paper gives out the result of a transient calculation with an advanced PWR model. Compared to some old integration method, the new method has far more strong stability. And the result shows that the integration progress of the code does not obviously affect the calculation accuracy, but definitely extends the application fields because of the multiple functions supplied by the RINSIM.
Evaluation of Explosivity From Fuel Coolant Interaction Experiments in the TROI Facility
Authors: Hong Seong-Wan , Park Rae-Joon 
Published: October 2018

The explosivity of corium is evaluated from the fuel coolant interaction experimental data produced in the TROI facility.
From 62 experimental datasets used to simulate the fuel coolant interaction under a partial flooded condition of the reactor, the explosivity when observing the dynamic pressure and/or force measured in experiments is qualitatively evaluated depending on the coolant depth, the shape of the interaction chamber, the free fall height of the melt jet, and the water temperature effect. For 12 of the experimental datasets produced under the OECD/SERENA Project, the explositivity is quantitavely evaluated using the conversion ratio. The conversion ratio is also evaluated for three experimental datasets to simulate the fuel coolant interaction under reactor flooded conditions.
The corium system showed a relatively low explosivity compared to the ZrO2 or Al2O3 system. It also turned out that the flooding condition of the reactor cavity does not affect the change in explosivity.
A Demonstration of Practical Elimination of Early or Large Fission Product Release for the UK ABWR Generic Design Assessment
Authors: Ang Ming Leang , Mohamud Nuh , Chitose Hiromasa , Hirokawa Naoki , Kimura Ryusuke 
Published: October 2018

The demand for continuous improvement in safety of nuclear power plants has led to an international expectation that early or large releases of fission products as a result of severe accidents be practically eliminated for new reactor designs. The UK Department for Business, Energy and Industrial Strategy has recently published the Seventh UK Report which provided confirmation of UK demonstrating compliance with the obligations of the Convention on Nuclear Safety [Ref-1]. Relating to Article 14 on the Assessment and Verification of Safety, the UK nuclear Regulator (Office for Nuclear Regulation (ONR)) has the expectation of the demonstration of ‘practical elimination’ of potential severe accident states be included in the safety cases for new nuclear power plants [Ref-2]. In order to achieve this, the safety case should show either that it is physically impossible for the accident states to occur or the states can be considered to be extremely unlikely with a high degree of confidence by design provisions [Ref-3] [Ref-4] [Ref-5]. A demonstration framework was developed and applied successfully in the UK ABWR Generic Design Assessment (GDA) Pre-Construction Safety Report (PCSR) which was submitted to the ONR in August 2017 [Ref-6].
A summary of this demonstration is provided in this paper.
Study on Flow Structure in a Supersonic Steam Injector
Authors: Kamata Yuki , Fujishiro Masaya , Kaneko Akiko , Abe Yutaka 
Published: October 2018

Steam injector (SI) are attracting attention as countermeasures against severe-accident in nuclear reactors. It is a static jet pump which operates using driving force to draw steam and water by internal pressure being reduced by direct contact condensation of these two fluids. In addition, capability of SI as a heat exchanger with high heat-transfer is expected. The absence of a drive unit such as an external power supply and rotating machine is significant characteristic of SI, and it can be expected to suppress the cost of installation and maintenance. It is also possible to produce a discharge pressure higher than the inlet pressure. From these facts, SI is expected to be applied as a static safety system that can cool the reactor core even if power lose at the nuclear power plant.
Although SI has been used for steam engines since long ago, the mechanism of its operation has not yet been clarified. Thus, elucidation of the mechanism of operation of SI is indispensable for introduction to a nuclear power plant. A one-dimensional analytical model which predicts the operating characteristics assuming full condensation and evaluated discharge pressure is constructed (Narabayashi et al., 1996). In addition, from detailed observation, it was confirmed by that there is a boundary of luminance in the diffuser section (Abe et al., 2012). This is considered as the boundary where the two-phase flow condenses. However, this phenomenon is not considered in the current analysis model.
The aim of this research is to clarify the flow structure in the diffuser section of SI. For that purpose, the state of the diffuser section of the transparent SI test part was observed with a highspeed camera, and the pressure at each point in it was measured simultaneously. The boundary of the luminance is confirmed to approach the throat as closing the back-pressure valve. In addition to this boundary, it was confirmed that the bright region intermittently propagated downstream. This phenomenon is supposed to be caused by pressure increasing, and the propagations assumed as a pressure wave moving at the sound speed. Thus, void fraction is estimated by calculating this propagation speed with image processing. Furthermore, experiments were carried out using three types of large, medium and small test parts, respectively. From the above results, the internal flow structure in the SI diffuser section was discussed.
Applying Computational Fluid Dynamics Methodologies to Hydro-Dynamic Loads in the Suppression Pool of a BWR
Authors: Mauritzson David , Perzon Sven , Probert Thomas 
Published: October 2018

The direct and indirect hydro-dynamic loads in the suppression pool of the BWR fleet were developed well before 2000. These loads were based on scaled experiments and numerical solutions using one dimensional models. The analysis was cumbersome and conservatisms were added in multiple steps.
As these loads spread throughout the containment, as global vibrations, they are generally a part of all structural verification inside the containment. This fact has made it hard and unpractical to challenge and revise these loads, as a change could lead to significant re-work. A consequence of this is that loads in the suppression pools have seldom been revisited, regardless if they cause local or global vibrations. This is problematic when new equipment is needed. The design of this equipment suffers from the very conservative loads.
Experience has shown that these loads can be challenged and refined using updated techniques, leading to significantly lower loads. This was realized during the modernization of Swedish plant Oskarshamn Unit 2 with Mark II containment, where loads following pool swell proved to be particularly challenging and it was decided to investigate possibilities to reduce the conservative loads.
Due to the large scale of the condensation pool coupled with the transient and small scaled condensation of steam at the drywell vent pipe nozzle full CFD resolution is not feasible. Instead lumped models in GOTHIC was used to increase resolution from the 1D approach of normal containment analysis, to a resolution that can account for the features of the condensation pool. This showed that the pool swell was less uniform than initially thought, leading to fewer objects affected as well as lower loads on objects that suffered from these loads. A full CFD analysis was then used to resolve phenomena working on even shorter time scales leading to a complete rework of all local loads.
The loads addressed using updated codes and modelling techniques was pool swell impact (PSI), pool swell drag (PSD) and local drag loads due to pressure relief valve opening, LV/SRV.
The current work shows that using updated modelling techniques and aligning results with previous analysis and documentation, it is possible to reduce loads for some events in the suppression pool without violating safety for the power plant. The results from the GOTHIC model shows that the cylindrical shape of the pool will create an uneven velocity distribution radially at pool swell resulting in much smaller loads at the outer boundary.
CFD analysis of the LV/SRV event shows that the loads are reduced in comparison with previous methodology and this is mainly because of shadow effects. The calculations also verifies assumptions used in the previous methodology.
Experimental Nusselt Number in Rod Bundles Cooled by Heavy-Liquid Metals
Authors: Pacio J. , Daubner M. , Wetzel T. , Di Piazza I. , Tarantino M. , Martelli D. , Angelucci M. 
Published: October 2018

A detailed safety assessment of innovative Generation IV reactor designs with heavy-liquid metal coolants, such as lead and lead-bismuth eutectic (LBE), requires an evaluation of the maximum core temperature in several postulated scenarios.
Considering the complex geometry of fuel assemblies (FAs), and the low Prandtl number of the coolants, this flow scenario is challenging for the models used in numerical simulations, e.g. for relating the turbulent transport of momentum and heat. Thus, reliable experimental data are needed for validation.
In recent years, a series of comprehensive heat transfer tests in fuel pin bundle simulators was performed at ENEA (Italy) and KIT (Germany) in the framework of the European collaborative projects THINS and SEARCH. Both grid and wire spacer geometries are considered, in a wide range of operating conditions (temperature, flow velocity and power density) representative of the ALFRED and MYRRHA fuel assemblies, in natural and forced-convective flow regimes. Although different experimental approaches were followed by each group (e.g. thermocouple position and average data treatment), there is a relatively good agreement on results in the overlapping regions.
These experimental studies indicate that the mean Nusselt number is in well agreement with the predictions of empirical correlations developed for sodium systems. In particular, for wire-spaced FA, heat transfer results show values close to the Kazimi-Carelli correlation both for low and high flow rates at ENEA and KIT respectively. For grid-spaced FA, results are more in agreement with Ushakov correlation.
Furthermore, large temperature differences are measured by thermo-couples installed at selected rods and sub-channels. A discussion on the influences of the spacer design and bundle size is included. This wide comparison allows an overview of the research on the HLM cooled fuel assembly in Europe.
Discussion About Sodium-Concrete Reaction in Presence of Internal Heater
Authors: Kawaguchi Munemichi , Miyahara Shinya , Uno Masayoshi 
Published: October 2018

Sodium-concrete reaction (SCR) is one of the important phenomena during severe accidents in sodium-cooled fast reactors (SFRs) owing to the generation of large sources of hydrogen and aerosols in the containment vessel [1–3]. In this study, SCR experiments were performed to investigate the chemical reaction beneath the internal heater (800 °C), which was used to simulate the obstacle and heating effect for SCR. Furthermore, the effects of the internal heater on the self-termination mechanism were discussed.
The internal heater on the concrete seemed to hinder the transport of Na into the concrete. Therefore, Na could react with the concrete at the periphery of the internal heater. The concrete ablation depth at the periphery was larger than under the internal heater. However, the Na concentration around the reaction front was about 30 wt.% despite the position of the internal heater. The Na concentration was similar to that of Na2SiO3, which was almost same as that in our past study [4–5]. It was found that the Na concentration condition was one of the dominant parameters for the self-termination of SCR, even in the presence of the internal heater.