scholarly journals CLEAR‐S: an integrated non‐nuclear test facility for China lead‐based research reactor

2016 ◽  
Vol 40 (14) ◽  
pp. 1951-1956 ◽  
Author(s):  
Y. Wu
Keyword(s):  
Research Reactor ◽  
Nuclear Test ◽  
Test Facility

Development and Preliminary Validation of a STH-CFD Coupling Method for Multiscale Thermal-Hydraulic Simulations of the MYRRHA Reactor

2016 ◽  
Author(s):  
A. Toti ◽  
J. Vierendeels ◽  
F. Belloni
Keyword(s):  
Research Reactor ◽  
Fluid Dynamic ◽  
Domain Decomposition Method ◽  
Test Facility ◽  
Research Centre ◽  
High Tech ◽  
Primary System ◽  
Complex Flow ◽  
Transient Event ◽  
Flow Mixing

MYRRHA (Multi-purpose hybrid research reactor for high-tech applications) is a lead-bismuth eutectic (LBE) cooled research reactor currently under development at SCK•CEN, the Belgian Nuclear Research Centre. The compact design of the pool-type primary system implies the presence of pronounced 3D thermal fluid-dynamic phenomena, which can affect the evolution of certain accidental transients such as loss of flow (LOF). System thermal-hydraulics (STH) codes, conceived to carry out global NPP safety analyses, present severe limitations in taking into account local 3D phenomena including flow mixing, thermal stratification, etc. To overcome this limitation, a promising solution is coupling STH codes with CFD codes, which can calculate complex flow fields but result, on the other hand, in too expensive computational resources for whole-plant simulations. A domain decomposition method that couples the STH code RELAP5-3D and the CFD code Ansys FLUENT has been developed and implemented. Proof-of-principle tests on simple configurations have been carried out to demonstrate its validity and to identify modeling and numerical issues. The experimental campaign carried out at the test facility TALL-3D, operated by the KTH Royal Institute of Technology in Sweden, has been selected for preliminary verification and validation (V&V) of this method. This paper presents the results of the coupled 1D-3D simulation of a forced-to-natural circulation transient event, whose evolution results to be strongly affected by flow mixing and stratification phenomena. The experimental validation, based on a high-quality set of experimental data, is currently on-going. Further development and validation activities will be carried out in the experimental facility ESCAPE, under commissioning at SCK•CEN, within the recently launched EU project MYRTE (Horizon 2020 programme).

scholarly journals Quality Management System Program of in Vitro/in Vivo Test Facility of Boron Neutron Capture Therapy at Kartini Research Reactor

2016 ◽  
Vol 1 (2) ◽  
pp. 108
Author(s):  
Widarto Widarto ◽  
Isman Mulyadi Tri Atmoko ◽  
Gede Sutresna Wijaya
Keyword(s):  
Boron Neutron Capture Therapy ◽  
Neutron Capture ◽  
Research Reactor ◽  
Test Facility ◽  
Neutron Capture Therapy ◽  
In Vivo Test ◽  
Boron Neutron Capture ◽  
System Program

The quality manajement system program of in vitro / in vivo test facility of  Boron Neutron Capture Therapy (BNCT) methode as quality assurance requirement for utilization of radial pearcing beamport of Kartini research have been done.  Identification and management of technical specification and parameters meassurement of to the radial piercing beamport have been determined for preparing in vitro / in vivo test facility. The parameters are epithermal neutron flux is  9,8243E+05  n cm<sup>-2</sup> s<sup>-1</sup>and  thermal neutron flux is 3,0691E+06 n cm<sup>-2</sup> s<sup>-1</sup>, radiation shielding of parafin,  dimension and size  of piercing radial and instrumentatin and control system for automatic transfer of in vitro / in vivo samplels have been documented. Management system of the documents for fullfil  basic guidance to perform working job of in vitro / in vivo at the piercing radial beamport of Kartini Research Reactor in order purpose utilization of the reactor  for safety worker of the radiation area, society  and invironment beeing safely

Mechanical and Thermal Assessment by BAM of a New Package Design for the Transport of SNF From a German Research Reactor

2020 ◽  
Author(s):  
Steffen Komann ◽  
Viktor Ballheimer ◽  
Thomas Quercetti ◽  
Robert Scheidemann ◽  
Frank Wille
Keyword(s):  
Research Reactor ◽  
Drop Test ◽  
Radioactive Material ◽  
Test Facility ◽  
Test Sequence ◽  
Thermal Safety ◽  
Package Design ◽  
Drop Tests ◽  
Approval Procedure ◽  
The Impact

Abstract For disposal of the research reactor of the Technical University Munich FRM II a new transport and storage cask design was under approval assessment by the German authorities on the basis of International Atomic Energy Agency (IAEA) requirements. The cask body is made of ductile cast iron and closed by two bolted lid systems with metal seals. The material of the lids is stainless steel. On each end of the cask the wood-filled impact limiters are installed to reduce impact loads to the cask under drop test conditions. In the cavity of the cask a basket for five spent fuel elements is arranged. This design has been assessed by the Bundesanstalt für Materialforschung und -prüfung (BAM) in view to the mechanical and thermal safety analyses, the activity release approaches, and subjects of quality assurance and surveillance for manufacturing and operation of the package. For the mechanical safety analyses of the package a combination of experimental testing and analytical/numerical calculations were applied. In total, four drop tests were carried out at the BAM large drop test facility. Two tests were carried out as a full IAEA drop test sequence consisting of a 9m drop test onto an unyielding target and a 1m puncture bar drop test. The other two drop tests were performed as single 9m drop tests and completed by additional analyses for considering the effects of an IAEA drop test sequence. The main objectives of the drop tests were the investigation of the integrity of the package and its safety against release of radioactive material as well as the test of the fastening system of the impact limiters. Furthermore, the acceleration and strain signals measured during the tests were used for the verification of finite-element (FE) models applied in the safety analysis of the package design. The FE models include the cask body, the lid system, the inventory and the impact limiters with the fastening system. In this context special attention was paid to the modeling of the encapsulated wood-filled impact limiters. Additional calculations by using the verified numerical model were done to investigate e.g. the brittle fracture of the cask body made of ductile cask iron within the package design approval procedure. The thermal safety assessment was based on analytical energy balance calculations and FE analyses. As an additional point of evaluation in frame of approval procedure, the effect of possible impact limiter burning under accident conditions of transport was considered by the applicant and assessed by BAM. This paper describes the package design assessment from the point of view of the competent authority BAM including the applied assessment strategy, the conducted drop tests and the additional calculations by using numerical and analytical methods.

Control Rod Withdrawal Analysis of the Supercritical Water Reactor-Fuel Qualification Test Facility in the LVR-15 Research Reactor

2015 ◽  
Vol 2 (1) ◽  
Author(s):  
Csaba Maráczy ◽  
György Hegyi ◽  
István Trosztel ◽  
Emese Temesvári
Keyword(s):  
Supercritical Water ◽  
Research Reactor ◽  
Three Dimensional ◽  
Reactor Fuel ◽  
Test Facility ◽  
Qualification Test ◽  
Control Rod ◽  
Water Reactor ◽  
Reactivity Coefficients ◽  
Supercritical Water Reactor

The aim of the supercritical water reactor-fuel qualification test (SCWR-FQT) Euratom-China collaborative project is to design an experimental facility for qualification of fuel for the supercritical water-cooled reactor. The facility is intended to be operated in the LVR-15 research reactor in the Czech Republic. The pressure tube of the FQT facility encloses four fuel rods that will operate in similar conditions to the evaporator of the HPLWR reactor. This article deals with the three-dimensional (3D) coupled neutronic-thermohydraulic steady-state and transient analysis of LVR-15 with the fueled loop. Conservatively calculated enveloping parameters (e.g., reactivity coefficients) were determined for the safety analysis. The control rod withdrawal analysis of the FQT facility with and without reactor SCRAM was carried out with the KIKO3D-ATHLET-coupled dynamic code.

scholarly journals Satellite radar observation of large surface collapses induced by the 2017 North Korea nuclear test

2020 ◽  
Vol 10 (1) ◽  
Author(s):  
Won-Kyung Baek ◽  
Hyung-Sup Jung ◽  
Tae Sung Kim
Keyword(s):  
North Korea ◽  
Surface Deformation ◽  
Complex Surface ◽  
Nuclear Test ◽  
Test Facility ◽  
Deformation Pattern ◽  
Vertical Deformation ◽  
Horizontal Deformation ◽  
Maximum Deformation ◽  
The North

Abstract The artificial earthquake of mb 6.1 related to the North Korea’s sixth nuclear test occured at Mt. Mantap, North Korea on September 3, 2017. It was reported that a large and complex surface deformation was caused by the event. The surface deformation was composed of expansion of explosions, collapse, compaction and landslides. Since the precise vertical deformation measurement is very important to estimate the stability of the nuclear test facility, we retrieved a precise 3D surface deformation field and then decomposed the vertical deformation pattern from the 3D deformation. The measured maximum deformation was about − 491, − 343 and 166 cm with the measurement uncertainty of about 3.3, 4.1 and 2.7 cm in the east, north and up directions, respectively. The maximum horizontal deformation was approximately 515 cm. The horizontal deformation clearly showed a radial pattern because it was mainly caused by the explosions and landslides, while the vertical deformation displayed a rugged pattern because it was affected by the explosions, compaction and collapse. The collapse may seem to occur along the underground tunnels and at the test site’s epicenter as well. Moreover, the severe collapse was observed westside from the epicenter of the sixth nuclear test, and it has a depth of about 68.6 cm on the area of 0.3765 km2. On the basis of our results including the shapes, locations and volume changes of the large collapse, evidently a new vital piece of information was obtained so that it could be used to interprete the sixth nuclear test more accurately.

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