Preliminary Design of Unloading Device for MNSR LEU Conversion

2018 ◽  
Author(s):  
Hao Qian ◽  
Li Yiguo ◽  
Peng Dan ◽  
Wu Xiaobo ◽  
Lu Jin ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Reactor Vessel ◽  
Cover Plate ◽  
Spent Fuel ◽  
Long Distance ◽  
Control Rod ◽  
Small Cover ◽  
Sealing Performance ◽  
Safety And Reliability ◽  
Application Requirements

In order to solve the problem that the current unloading operation will destroy the sealing performance of Miniature Neutron Source Reactor (MNSR) reactor vessel and the tightness can’t be restored, and to meet the application requirements that the original reactor vessel will be reloaded and operated after MNSR LEU conversion, the new unloading device is designed, which can be used without separation of reactor vessel. There has only one fuel assembly in MNSR. When the fuel assembly are unload for MNSR LEU conversion, the cover plate of the pool is removed, the cadmium string is put in, and the neutron detector is placed at first. After removing the drive mechanism and the control rod, and opening the small cover plate at the top of reactor vessel, the fuel assembly can be grabbed and unloaded by unloading tool only through the opening of the small top cover plate. The MNSR spent fuel has very high radioactivity. The auxiliary mechanical device can be used with unloading tools to realize operation in a long distance by lifting and level motion, which is convenient to shield and can reduce the works’ irradiation dose level effectively. Through calculation and analysis, the results show that the structure strength of unloading device is much larger than the actual load to ensure operation safety and reliability. The unloading device is easy to process and operate, and can be used in the practical operation of MNSR LEU conversion or decommissioning at home and abroad to simplify the operation steps and improve the working efficiency.

Insulation Material Deposition and Distribution in a PWR Fuel Assembly Cluster

2012 ◽  
Author(s):  
Stefan Renger ◽  
Sören Alt ◽  
Wolfgang Kästner ◽  
André Seeliger ◽  
Frank Zacharias
Keyword(s):  
Fuel Assembly ◽  
Reactor Core ◽  
Visual Observation ◽  
Test Facility ◽  
Insulation Material ◽  
Control Rod ◽  
Free Jet ◽  
Laser Sensors ◽  
Material Deposition ◽  
Loss Of Coolant Accident

Background of experimental and methodical work is the loss of coolant accident (LOCA) with release of fibrous pipe insulation material. Latest investigations were focused on material deposition and distribution (cross mixing) in the reactor core. Therefore, a 2×2 PWR fuel assembly (FA) cluster was constructed. Four shortened PWR-FA-dummies are provided with separated in- and outlets. Every 16×16 fuel rod dummy consists of 20 control rod simulators, two spacers, FA-head and FA-bottom with a 3.5×3.5 mm integrated debris-screen filter (IDF). The cluster is encased in an acrylic housing for visual observation. It is connected with the test facility “Zittau Flow Tray” (ZFT), a simplified sump model, which allows inclusion and investigation of complex phenomena like material sedimentation in the sump and strainer blockages. A well mixing of air in the fluid was also considered by free jet expansions and flows through full cone-nozzles as well as marginal air entrainments. This Paper includes descriptions of applied measuring techniques (digital image processing, thrubeam laser sensors etc.) and an overview of all considered boundary conditions. Experimental results, aiming at the development, implementation and verification of multiphase flow and strainer models, are presented.

Segmentation and Removal of the Carolinas-Virginia Tube Reactor (CVTR) Moderator Tank

2007 ◽  
Author(s):  
Michael G. Anderson
Keyword(s):  
South Carolina ◽  
Reactor Vessel ◽  
Spent Fuel ◽  
Tube Reactor ◽  
Activation Products ◽  
Reactor Building ◽  
Spent Fuel Pool ◽  
Steel Liner ◽  
Water Cutting ◽  
Thermal Shields

Special tooling has been deployed to segment the Moderator Tank (MT) at the Carolinas-Virginia Tube Reactor (CVTR) Parr site near Jenkinsville, South Carolina. The MT or reactor vessel, the most activated component remaining on site which included over 1,000 Ci of activation products, has been segmented into sections to fit within three hardware liners and three custom boxes. This work has been completed in approximately 12 months from tool conception to final packaging with no spread of contamination, no generation of secondary wastes and minimizing personnel radiological exposure. With contact dose readings in excess of 90 R/hr, segmentation of the MT had to be performed remotely and with the assurance that the spread of contamination to otherwise clean areas of the reactor building did not occur. Additionally, since the MT was entombed within a bioshield not capable of containing water, cutting had to be performed dry without benefit of shielding typically provided by the water of a spent fuel pool. In addition, the component removal scope included the removal, packaging and disposal of other activated components including thermal shields and the steel liner from the internal face of the bioshield. Concept engineering began in January 2006. Tools were tested and delivered in May 2006. Segmentation was completed in December 2006, followed by the removal of the thermal shields and bioshield liner. The component removal work was completed without the spread of contamination, no generation of secondary waste and an exposure total of 17 person rem.

A Review of Alloy 600 Cracking in Operating Nuclear Plants Including Alloy 82 and 182 Weld Behavior

2004 ◽  
Author(s):  
Warren Bamford ◽  
John Hall
Keyword(s):  
Weld Metal ◽  
Base Metal ◽  
Steam Generator ◽  
Structural Integrity ◽  
Reactor Vessel ◽  
Alloy 600 ◽  
Control Rod ◽  
Nuclear Plants ◽  
Steam Generator Tubing ◽  
Long Time

Service induced cracking in Alloy 600 has been known for a long time, having been first observed in the 1980’s in steam generator tubing and small bore piping, and later, in 1991, in reactor vessel control rod drive mechanism (CRDM) head penetrations. Other than steam generator tubing, which cracked within a few years of operation, the first Alloy 600 cracking was in base metal of Combustion Engineering small bore piping, followed closely by CE pressurizer heater sleeves. The first reactor vessel CRDM penetrations (base metal) to crack were in France, US plants found CRDM cracking several years later. Three plants have discovered weld metal cracking at the outlet nozzle to pipe weld region. This was the first known weld metal cracking. This paper will chronicle the development of service-induced cracking in these components, and compare the behavior of welds as opposed to base metal, from the standpoint of time to crack initiation, growth rate of cracks, and their impact on structural integrity. In addition, a discussion of potential future trends will be provided.

Industrial Remote Control System Based on C8051F120

2015 ◽  
Vol 743 ◽  
pp. 176-179
Author(s):  
Hong He ◽  
Xiao Jun Xu ◽  
Zhi Hong Zhang ◽  
Cheng Qian Mao
Keyword(s):  
Control System ◽  
Remote Control ◽  
Communication Theory ◽  
Wireless Transmission ◽  
Climbing Robot ◽  
Single Chip ◽  
Long Distance ◽  
Remote Control System ◽  
Potential Safety ◽  
Safety And Reliability

In order to solve the potential safety risk of traditional industrial remote control, a deep research on embedded and wireless communication theory has been done and a wireless remote control system with the C8051F120 single chip as the main control chip has been invited. Data can be sent and received steadily between transmitting terminal and receiving terminal in this system even in unfavorable environment, which largely improves the safety and reliability for long-distance operation. This system, with the feature of small size, easy installment, low power consumption and excellent wireless transmission, can be applied to industrial machinery such as pump car and wall-climbing robot.

Effect of CRGT Cooling on Modes of Global Vessel Failure of a BWR Lower Head

2012 ◽  
Author(s):  
Walter Villanueva ◽  
Chi-Thanh Tran ◽  
Pavel Kudinov
Keyword(s):  
Heat Transfer ◽  
Vessel Wall ◽  
Reactor Vessel ◽  
Creep Model ◽  
Guide Tube ◽  
Mechanical Loads ◽  
Control Rod ◽  
Melt Pool ◽  
Creep Analysis ◽  
Lower Head

An in-vessel stage of a severe core melt accident in a Nordic type Boiling Water Reactor (BWR) is considered wherein a decay-heated pool of corium melt inflicts thermal and mechanical loads on the lower-head vessel wall. This process induces creep leading to a mechanical failure of the reactor vessel wall. The focus of this study is to investigate the effect of Control Rod Guide Tube (CRGT) and top cooling on the modes of global vessel failure of the lower head. A coupled thermo-mechanical creep analysis of the lower head is performed and cases with and without CRGT and top cooling are compared. The debris bed heat-up, re-melting, melt pool formation, and heat transfer are calculated using the Phase-change Effective Convectivity Model and transient heat transfer characteristics are provided for thermo-mechanical strength calculations. The creep analysis is performed with the modified time hardening creep model and both thermal and integral mechanical loads on the reactor vessel wall are taken into account. Known material properties of the reactor vessel as a function of temperature, including the creep curves, are used as an input data for the creep analysis. It is found that a global vessel failure is imminent regardless of activation of CRGT and top cooling. However, if CRGT and top cooling is activated, the mode and timing of failure is different compared to the case with no CRGT and top cooling. More specifically, with CRGT and top cooling, there are two modes of global vessel failure depending on the size of the melt pool: (a) ‘ballooning’ of the vessel bottom for smaller pools, and (b) ‘localized creep’ concentrated within the vicinity of the top surface of the melt pool for larger pools. Without CRGT and top cooling, only a ballooning mode of global vessel failure is observed. Furthermore, a considerable delay (about 1.4 h) on the global vessel failure is observed for the roughly 30-ton debris case if CRGT and top cooling is implemented. For a much larger pool (roughly 200-ton debris), no significant delay on the global vessel failure is observed when CRGT and top cooling is implemented, however, the liquid melt fraction and melt superheat are considerably higher in non-cooling case.

scholarly journals BOUNDARY CONDITION MODELING EFFECT ON THE SPENT FUEL CHARACTERIZATION AND FINAL DECAY HEAT PREDICTION FROM A PWR ASSEMBLY

2021 ◽  
Vol 247 ◽  
pp. 12008
Author(s):  
Augusto Hernandez-Solis ◽  
Klemen Ambrožič ◽  
Dušan Čalič ◽  
Luca Fiorito ◽  
Bor Kos ◽  
...  
Keyword(s):  
Boundary Condition ◽  
Fuel Assembly ◽  
Emission Rate ◽  
Neutron Emission ◽  
Spent Fuel ◽  
Decay Heat ◽  
Fuel Pin ◽  
Condition Modeling ◽  
Fuel Characterization ◽  
Relative Differences

In this paper, two main exercises have been carried out to describe the effect that varying an albedo boundary condition has in the computation of observables such as decay heat, neutron emission rate and nuclide inventory from a PWR fuel assembly (or a configuration of assemblies) during a depletion scenario. The SERPENT2 code was then employed to emphasize the importance of modeling a proper boundary condition for such purposes. Moreover, the effect of taking into account more than a single fuel-pin region for depletion studies while varying the type of boundary condition, was also accounted for. The first exercise has the main objective of comparing in a single fuel assembly the albedo variations ranging from 1.1 up to full vacuum conditions. By comparing to the reference assembly (considered to be the case of full reflective conditions), relative differences up to +17% were observed in decay heat and up to almost -30% in neutron emissions. Also, a clear dependence on the albedo was detected if more than one depletable zone was considered while computing the integral value of observables of interest. Regarding the second exercise, where a 3 × 3 configuration of fuel assemblies is being now considered with a reflector section in the middle, a negligible effect on the observables was observed for the single fuel pin zone case; instead, an effect in the 244Cm computation when analyzing two fuel pin-zones produced a change in the neutron emission rate during cooling time up to 2.5% (while comparing it to the reference single assembly case).

Using RIAM for Optimizing Reactor Vessel Head Leak Failure Mode Maintenance Strategies

2004 ◽  
Author(s):  
Vera Erguina ◽  
Ernest J. Kee
Keyword(s):  
Failure Mode ◽  
Cost Benefit ◽  
Leak Detection ◽  
Reactor Vessel ◽  
Failure Rates ◽  
Maintenance Strategies ◽  
Control Rod ◽  
Control And Prevention ◽  
Plant Valuation

A cost-benefit-risk study of the best (in terms of plant valuation) long-term strategy for canopy seal weld leakage control and prevention including a span of options for reactor vessel head J-Groove weld failures is performed. The range of options goes from the “do nothing” option up to replacement of the reactor vessel head with new control rod drive mechanisms. The results are presented in form of net present values for the options cost and include uncertainties for long-term costs associated with the options activities and prediction of failure rates for control rod drive mechanisms canopy seals and J-Groove welds. Results of this study rely on catching leaks during an outage and not during operation (mid-cycle). The consequence associated with a mid-cycle leak is much higher than for one found during an outage. This tends to indicate rigorous visual inspections during outages to ensure repairs are made during refueling outages. Thus, the effect of forced outage due to mid-cycle leak detection is not included to avoid strongly biasing the results in favor of reactor vessel head replacement.

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