bluuebunny/crossref_metadata_2025_split huggingface dataset:
Research on Simplified Method of Transients Combining and Loading in Fatigue Crack Growth Analysis of Carbon Steel Nuclear Piping
Authors: Liu Zhenshun , Zhen Hongdong 
Published: October 2018

The fatigue crack growth analysis of nuclear piping is a nonlinear calculation process. The loading sequence and combination mode could affect the results. How to consider these effects is unclear. Fatigue crack growth analysis includes a large number of nonlinear iterative calculations, and the calculation speed is slow.
This paper selects carbon steel nuclear piping as the research object. Based on the analysis process provided by ASME code XI volume, a simplified analysis method is explored by introducing the reference crack depth a’ and the envelope transient. The simplified analysis method is conservative because it has been proved that the crack growth rate is positively related to the crack size only if the maximum stress intensity factor is greater than 0 and the minimum stress intensity factor is less than 0. The simplified analysis method is proved to be able to significantly improve the calculation speed by comparing the number of iterative calculations in the simplified analysis method and in the conventional analysis method.
The results indicate that the simplified analysis method could provide a conservative way of loading and combining the complex transients and could significantly reduce the number of nonlinear iterative calculations in the process of crack fatigue growth analysis for carbon steel nuclear piping when the maximum stress intensity factor greater than 0 and the minimum stress intensity factor is less than 0.
A Critical Experimental Study of Bubble Effect in the Process of Spent Fuel Dissolving
Authors: Xiaoping Zhou , Shuhong Liang , Zhaodong Xia 
Published: October 2018

The influence of gas introduction on the critical safety of the nuclear fuel system under the condition of cold condition, given reactor material and geometry structure is studied.
Refer to bubble effect test experiment on nuclear critical safety test device (YSR) and considering solid-liquid two-phase nuclear fuel system with uranyl nitrate solution - uranium dioxide fuel element as the experimental platform, the dynamic process of the real behavior of bubbles in uranyl nitrate solution has been simulated in the quasi-static way by replacing bubble generator with aluminous bubble simulation elements.
Bubble effect is the reactivity change caused by the change of volume of solution, neutron leakage and absorption property in the nuclear fuel system due to the bubbles generated in the solution. In the dissolving process of spent fuel, oxygen or nitrogen are usually added to accelerate the dissolution of fuel element shear section, and some other bubble production are also caused by the heat released during the dissolution process. Here, the bubble production caused by the heat is omitted and only artificial gas introduction is considered in my study. When there are enough bubbles in the uranium solution system, the volume of the solution will increase, which will inevitably influence the absorption and leakage property of the neutrons, and further influence the reactivity of the nuclear fuel system.
The corresponding relationship between the bubble-intake rate and the bubble equivalent diameter, arising velocity and bubble share is determined through fluid dynamics modeling to manufacture the aluminous bubble simulation elements. The theoretical calculation by MONK9A and the critical experimental measurements are also compared and analyzed in this paper.
The results showed that the reactivity caused by bubbles was negative, and the greater the intake rate, the greater the negative effect. Meanwhile the theoretical calculated value was in good agreement with the experimental value and the maximum deviation was 63.4 pcm.
Experimental Study on Flow Instability During Gravity-Driven Reflooding
Authors: Tong Lili , Zhang Dandi , Hou Liqiang , Wang Xiaoji 
Published: October 2018

Due to the inherent safety characteristics of passive core injecting system, it is one of the most important mitigation strategies under loss of coolant accident (LOCA). However, flow instability during passive gravity-driven reflooding may occur, which can obstruct core reflooding process and reduce the heat removal rate from the reactor core. Therefore, exploring the characteristic and influencing mechanisms of flow instability during gravity-driven reflooding is necessary to analyze the effectiveness of passive core injecting system and mitigate the consequences of LOCA.
In this work, small scale experiments are performed to investigate the flow instability during gravity-driven reflooding when the open ratio of steam outlet are in the ranges of 3.33–50%, gravity driving head are 16–20kPa and system pressure are 0.1–0.3MPa. The test facility consists of a cooling water storage tank, a test section and a condensate tank. The results show that flow instability could occur under specific conditions and have a strong regularity, which is divided into three stages: cooling water initial injection, cooling water expulsion and cooling water re-injection. In addition, the effects of open ratio of steam outlet, gravity driving head and system pressure on flow instability phenomenon are investigated. Increasing the open ratio of steam outlet accelerates the discharge rate of steam and prevents the accumulation of steam at the inner cylinder, which can contribute to cooling the heated rods and restrict the flow instability. And under low gravity driving head of 16kPa and low system pressure of 0.1MPa the temperature and pressure acutely change and flow instability is more likely to occur.
Computational Fluid Dynamics As a Tool for Deriving Subchannel Model Parameters: The PSBT Case Study
Authors: Puragliesi R. , Mukin R. , Clifford I. , Ferroukhi H. , Seidl M. 
Published: October 2018

The development of subchannel models for fuel assemblies and reactor cores requires accurate information on flow distribution, wall friction and loss coefficients in order to accurately predict the pressure, temperature and flow distribution on a subchannel level. This paper discusses the use of Computational Fluid Dynamics (CFD) simulations as a practical tool for characterising inlet velocity boundary conditions, an approximation of wall friction factor and spacer grid pressure loss coefficient, which are of fundamental importance to correctly generate a consistent subchannel model of a given assembly system. The geometry of the simplified PWR assembly presented here is based on the NUPEC PWR subchannel and bundle tests. Comparison of the derived friction factors and grid pressure loss coefficient with published and recommended values are reported. Discrepancies are also explained using additional calculations. A comparison of the overall system pressure drop, by comparing_numerical and analytical solutions, and the local axial pressure distribution at subchannel level are presented. To make a one-to-one comparison between CFD and subchannel solutions, volume-averaging is applied to the CFD results according to the chosen subchannel nodalization.
The obtained results show a perfect agreement between the two codes. This outcome reflects the correct approach employed to build two consistent numerical models by properly carrying important information from the high-resolution models (CFD) to the low-resolution models (subchannel code). Furthermore, it has been found that the large discrepancies recorded in the CFD prediction of the grid pressure loss coefficients suggested in the benchmark specifications are mainly because the suggested benchmark values do not take into account the presence of the bounding channel that was present in the experimental facility.
Design for Plant Modularisation: Nuclear and SMR
Authors: Wrigley Paul , Wood Paul , Stewart Paul , Hall Richard , Robertson Dan 
Published: October 2018

The UK Small Modular Reactor (UKSMR) programme has been established to develop an SMR for the UK energy market. Developing an SMR is a multi-disciplinary technical challenge, involving nuclear physics, electrical, mechanical, design, management, safety, testing to name but a few.
In 2016 Upadhyay & Jain performed a literature review on modularity in Nuclear Power. They concluded that although modularisation has been utilised in nuclear to reduce costs, more work needs to be done to “create effective modules”. Hohmann et al also concluded the same for defining modules in the chemical process plant industry.
The aim of this paper is to further define modules with a particular focus on an SMR for the UK market, the UKSMR. The methods highlighted may be relevant and applied to other international SMR designs or other types of plant.
An overview and examination of modularisation work in nuclear to date is provided. The different configurations are defined for the Nuclear Steam Supply System (NSSS) in primary circuits and then for Balance of Plant (BOP) modules. A top level design process has been defined to aid in the understanding of design choices for current reactors and to further assist designing balance of plant modules.
The paper then highlights areas for additional research that may further support module design and definition.
The Influence of Centrifugal Pump Characteristics on Dynamic Loadings on Pipelines After Power Failure
Authors: Marcinkiewicz Jerzy , Karaskiewicz Krzysztof , Joheman Claes 
Published: October 2018

The paper presents a Relap5 study of the influence of the centrifugal pump characteristics on the dynamic loads on piping system after power failure. Interpolated and experimental pump characteristics are used.
The differences between the interpolated and measured pump curves, the general description of Relap5 model and results of calculations in form of selected time curves for rotational speed, volume flow, pressures and dynamic forces are presented and discussed.
The analysis of the results shows that the maximal dynamic force on pipe section calculated with experimental pump curves can be up to 6 % higher than respective calculated using interpolated curves. However, it is not possible to determine to what extent the differences are caused by the interpolation itself or caused by the differences in the design of the centrifugal pumps. The latter since it differs more than 50 years between the pumps whose characteristics are used for interpolation and the pumps with corresponding experimental characteristics.
Transient Reactor Test Facility Restart 23 Years Later
Authors: Heath Bradley K. , Race Cody C. , Nelson Lee O. 
Published: October 2018

The Transient Reactor Test (TREAT) Facility, located at the Idaho National Laboratory (INL), is a versatile test facility able to subject experimental specimens to various transient nuclear conditions. TREAT was placed in standby after operating from February 1959 through April 1994, resulting in the loss of nearly all transient testing capability in the United States. Recently, the US Department of Energy (DOE) determined this capability was again needed. After DOE completed National Environmental Policy Act actions in February 2014, INL established the Resumption of Transient Testing Program (RTTP). RTTP was a multi-year effort to restart TREAT to reestablish a domestic transient testing capability. After 23 years of standby operations, the RTTP completed restart activities on August 31, 2017, 13 months ahead of schedule and nearly $20 million under budget. RTTP activities included an Environmental Assessment that resulted in a Finding of “No Significant Impact” associated with restarting TREAT, establishment of a compliant Safety Analysis Report (SAR), refurbishment and/or replacement of key reactor systems and components, key system knowledge recovery, reestablishment of configuration management, procedure updates, personnel training and qualification, and demonstration of operational readiness for reactor operations. Several noteworthy factors that contributed to the restart of TREAT include:
• Funding to acquire personnel and material resources provided in a timely fashion.
• Close coordination with the regulator’s (DOE) nuclear safety program during updates, interactive review, and approval of safety documentation provided for timely update of the TREAT SAR and implementing documents.
• Effective management control enabled by utilization of standard outage management techniques with a focus on age-related degradation and updated standards and requirements.
• DOE program management ensured efficient implementation of program management tools. These tools focused on clear high-level milestones and spend plans allowing flexibility for the contractor to respond to evolving facility conditions and information in a near-real time manner and with minimal program overhead. This approach enabled efficient execution of work in an environment where determination of required work scope was dependent on performance of inspection, testing, analysis, and evaluation activities.
• Implementation of the Contractor Assurance System, with frequent internal and externally-led assessments that facilitated process improvements and corrective actions to ensure the operational readiness for required contractor and DOE readiness assessments and safe nuclear operations.
• The RTTP benefited from archived plant documentation and maintenance performed while the plant was in a safe-standby status.
• Unique methods of reactivity control allowed for individual and integrated reactor system functional testing, procedure vetting, and personnel training while maintaining the reactor in a safe state.
Feasibility Assessment of Air-Cooling System As an Ultimate Heat Sink of the ATOM System
Authors: Shin Doyoung , Seo Gwang Hyeok , Na Min Wook , Kim Sung Joong , Kim Yonghee , Lee Jeongik 
Published: October 2018

Nowadays Small Modular Reactors (SMRs) have been receiving considerable attentions worldwide for potential advantages of an excellent flexibility for siting, low capital investment, and advanced safety. In Korea, a new research project has launched for the development of a conceptual design of a further advanced SMR which aims for a naturally-safe and autonomous operation, so called Autonomous Transportable On-demand reactor Module (ATOM). Major design objectives of the ATOM system are focused on the soluble boron-free (SBF) primary coolant system which enables the SMR to operate automatically in a load following mode. For the secondary system, the SCO2 power conversion cycle with air-cooling system as a final heat sink is being considered. The air-cooling system is expected to show flexible response even to extreme environmental conditions, such as a desert where utilization of cooling water is limited.
The objective of this study is a feasibility assessment for applying the air-cooling system as a final heat sink of the ATOM by means of experimental work. As a 1st phase of the ATOM development, we first conducted the experiments using a typically considered primary coolant, water-steam, to verify that air flow has enough cooling capability to remove developed heat which the coolant carries. An Integrated Condensation Loop with Air-cooling System (ICLASS) experimental facility with three pressure boundaries (Steam, coolant, and air) was established. The cooling capability of the air-cooling system was evaluated by varying steam mass flow rate, coolant flow rate, and air environment temperature as experiment variables. Overall heat transfer rate by condensation was compared with numerical simulations of a 1D thermal-hydraulics analysis code, using the MARS model of the ICLASS facility.
Low-Cycle Strength of Elements of Constructions
Authors: Zvorykin Alexander , Popov Roman , Bobyr Mykola , Pioro Igor 
Published: October 2018

Analysis of engineering approach to the operational life forecasting for constructional elements with respect to the low-cycle fatigue is carried out. Applicability limits for a hypothesis on existence of generalized cyclic-deforming diagram in case of complex low-cycle loading (deforming) are shown.
It is determined, that under condition of plane-stress state and piecewise-broken trajectories of cycle loading with stresses and deformation checking the cyclic deforming diagram is united in limits of deformations, which are not exceeded 10 values of deformation corresponding material yield point. Generalized kinematic equation of material damageability is described.
The method of damageability parameter utilization for increasing of accuracy calculation of structural elements low-cycle fatigue by using the effective coefficients of stresses and deformations taking into account the damageability parameter is given.
Reactor Core Cooling Performance of a Passive Endothermic Reaction Cooling System During Design and Non-Design Basis Accidents
Authors: Murray Nathan R. , Sailsbery Mitchell E. , Bischoff Samuel E. , Wilding Paul R. , Memmott Matthew J. 
Published: October 2018

A passive endothermic reaction cooling system (PERCS) is proposed to provide reactor core cooling during a station blackout (SBO). During a SBO, a PWR in which PERCS has been installed has a peak reactor core outlet temperature remains below 640 K (692.3°F) for 30 days, which is well below the nominal accident core outlet temperature during a SBO. During a LOCA, LOFA, and LOHSA, installation of a PERCS has no significant impact on safety performance. It should be noted that the PERCS will represent a minimal heat source (unless the PERCS is very large) during DBAs as emergency systems lower the coolant temperature below the PERCS temperature.
A typical PWR with an installed PERCS is modeled using RELAP5-3D. The results of the model demonstrate the high level of passive safety afforded by the PERCS which contributes to the mitigation of SBO consequences without adversely affecting nuclear plant safety during a LOCA, LOHSA, or LOFA. Future work in validating the PERCS as a method of passive safety for existing light water reactors is underway, including the refining the physical design, determining better kinetic and thermodynamic properties for MgCO3, updating the PERCS model, and using a more robust PWR plant model.
What Is a “Known and Established” Standard?
Authors: Miranda Samuel 
Published: October 2018

The NRC staff, and licensees will sometimes cite a Known and Established Standard, especially when there is evidence that the staff’s position, with respect to a Known and Established Standard, may have changed. Licensees will often flag such changes as backfits, and demand that they meet the requirements of the Backfit Rule in the Code of Federal Regulations (10 CFR §50.109). However, the term, Known and Established Standard, is not clearly defined. If the term is not defined by the NRC staff, then it’s possible that it will be defined by others, most notably by licensees with technical issues or license amendment requests (LARs) that are undergoing the staff’s review.
What follows is a discussion of the origin and application of the NRC’s Known and Established Standard, and how it might be defined. The Known and Established Standard is first deconstructed into its two components, Known and Established, and then each is defined separately. When this is done, it becomes apparent that some Known and Established Standards can be more important than others. The two definitions can be used, together, to distinguish between unequal, even incompatible Known and Established Standards. This could help the NRC staff identify the most relevant Known and Established Standard to apply in its review or evaluation of a particular technical issue or license amendment request (LAR). Simply choosing the most recent of Known and Established Standards might not be adequate.
QUENCH-06 Experiment Post-Test Calculations and Integrated Uncertainty Analysis With RELAP/SCDAPSIM/MOD3.4 and MOD3.5
Authors: Allison C. , Le B. T. , Gerova G. , Spasov I. , Perez-Ferragut M. , Hohorst J. 
Published: October 2018

The Quench-06 experiment (ISP-45) has been used as a benchmark and training aid for Innovative Systems Software (ISS) and our users/students since it was completed in the early 2000s. The experiment was first analyzed by several international organizations using RELAP/SCDAPSIM/MOD3.2. These results were submitted to the “blind” and “open” phases of the ISP. The experiment was subsequently used for basic user training for experimental analysis by our RELAP/SCDAPSIM/MOD3.4 and MOD3.5 users. It is also used extensively in our university support and training internships.
This paper describes an integrated uncertainty analysis of the QUENCH-06 electrically heated experiment, looking at the influence of uncertainties in experimental conditions and important models/correlations. The QUENCH calculations demonstrated the use of the new IUA, “Integrated Uncertainty Analysis”, option introduced into RELAP/SCDAPSIM/MOD3.4 in the summer of 2017 and MOD3.5 in the fall of 2017. The input models and results from both versions are discussed in the paper. The MOD3.4 results are based on the original input model developed for MOD3.2 and refined in the open phase of the ISP. The MOD3.5 results are based upon two base input models. The first was developed specifically to test the impact of MOD3.5 modeling improvements for the Quench electrically heated fuel rod simulator and was used in an early paper presented at this meeting in 2014. The second input model has been refined as part of the university support and training internship program and was used originally in 2016 to look at the influence of different approaches in modeling the insulated shroud used to minimize radial heat losses.
The uncertainty analysis provided in this paper looks at the influence of uncertainties in (a) the parabolic equations for Zircaloy oxidation, (b) the tungsten heater element resistances, (c) the convective heat transfer coefficients, (d) the contact resistance of the heater elements, and (e) the thermal conductivity of the porous zirconia used in the shroud. The uncertainty analysis demonstrated very clearly that a bias was introduced into the 2014 MOD3.5 input model. This bias was subsequently determined to be primarily associated to the modeling of the shroud and associated radial heat losses. This bias was reduced in the 2016 version of the MOD3.5 input model and will be further refined as our training activities on the modeling of experiments continue.
Upgrade of Accident Sequence Model of PRA for LOCA Inside PCV
Authors: Yamagishi Shohei , Tanno Shunsuke , Sato Teruyoshi , Saito Toshiteru , Hiraide Masayuki , Kita Toshinobu 
Published: October 2018

In Japan, Risk-informed activity using PRA is becoming increasingly active and Reactor Oversight Process (ROP) will be introduced in 2020. Then, PRA has attracted more attention, and it is necessary to immediately develop its infrastructures.
Therefore, as the first step, TEPSYS/TEPCO conduct to upgrade the internal event level 1 PRA that is the basis of all events targeting unit 7 (ABWR) of TEPCO Kashiwazaki-Kariwa Nuclear Power Plant. In particular, LOCA inside PCV for all piping connecting RPV including instrumentation and control piping reclassified based on its break size, break location, success or failure of the vapor suppression function, the accident progression during mission time, and PCV control procedures.
Especially, the finding presented in this paper address the newly grouping that break location is classified into not 2 (steam phase/liquid phase) groups [1] but 3 groups as below.
✓ RPV nozzles are located above L8 reactor water level (steam phase)
✓ RPV nozzles are located between TAF and L8 reactor water level (steam phase / liquid phase)
✓ RPV nozzles are located below TAF reactor water level (liquid phase)
In the Results, completeness of the accident sequence for LOCA inside PCV was improved to be applied to risk-informed activity and the proactive safety review.