Development of external coupling for calculation of the control rod worth in terms of burn-up for a WWER-1000 nuclear reactor

The new EPR™ Evolutionary high-Power nuclear Reactor is based on the best design concepts of French and German experiences over the past thirty years in power plant design construction and operation. In the framework of detailed engineering phase of plant construction in Finland (Olkiluoto 3) and in France (Flamanville 3), various investigations were led to improve the knowledge of hydraulics in innovative Reactor Pressure Vessel internals. A specific attention has been paid to hydraulics in Control Rod Guide Assembly (CRGA), which is an innovative component derived from the KONVOI design but adapted to the EPR™ design. In order to assess the hydraulic behavior of the EPR™ CRGA, a complementary experimental approach was developed by EDF and AREVA. Two mock-ups, representative of the full-scale EPR™ Control Rod Guide Assembly, were designed: MAGALY mock-up using real prototype of CRGA within AREVA’s engineering teams and PHACETIE mock-up at same scale but dedicated to a more analytic approach designed by EDF. The aim of the paper is to present the tests performed on two mock-ups to characterize the hydraulic behavior of this innovative EPR™ CRGA. At first, a comparison of the two mock-ups was done. The benchmark showed that the two mock-ups feature similar hydraulic behavior. The head losses evaluated along the experimental devices are practically identical. Moreover, the results showed that the global hydraulic forces applied to the Rod Cluster Control Assembly (RCCA) just up to the Upper Core Plate (UCP) are in a very good agreement. The good comparison between the two mock-ups strengthens the hydraulic characterization methods of the hydraulics in EPR™ CRGA. Specific studies were also led on the two mock-ups. The MAGALY mock-up was used to characterize the impact of the EPR™ operating conditions on the CRGA hydraulics behavior and especially on the vibrations and the drag forces applied to RCCA. Complementary measurements by using Laser Velocimetry techniques (PIV and LDV) were carried out on the PHACETIE mock-up allowing an evaluation of the velocity field inside the CRGA. The results obtained thanks to the two mock-ups allowed AREVA and EDF to have a good knowledge of the hydraulics behavior in the EPR™ CRGA. In a next step, further studies will be performed in order to deepen the knowledge of the vibratory behavior of the RCCA and the control rod drop time.

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Control rod worth calculation for VVER-1000 nuclear reactor using WIMS and CITATION codes

Progress in Nuclear Energy ◽

10.1016/j.pnucene.2008.03.003 ◽

2009 ◽

Vol 51 (1) ◽

pp. 184-191 ◽

Cited By ~ 16

Author(s):

Amir Hosein Fadaei ◽

Saeed Setayeshi

Keyword(s):

Nuclear Reactor ◽

Control Rod

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SAFETY ANALYSIS OF NEUTRON INTERACTION WITH MATERIAL PRACTICUM MODULE FOR THE KARTINI INTERNET REACTOR LABORATORY

JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA ◽

10.17146/tdm.2020.22.3.6017 ◽

2020 ◽

Vol 22 (3) ◽

pp. 104

Author(s):

Prasetyo Haryo Sadewo ◽

Puradwi Ismu Wahyono

Keyword(s):

Radiation Protection ◽

Nuclear Reactor ◽

Power Level ◽

Safety Analysis ◽

Reactor Physics ◽

Control Rod ◽

Reactor Kinetics ◽

Additional Equipment ◽

Activation Products ◽

And Control

Kartini Research Reactor, which is situated in Yogyakarta, is a 100 kW TRIGA (Training, Research, and Isotope Production by General Atomic)-type reactor mainly used for educational and training purposes. A system for remote learning on nuclear reactor physics named the Internet Rector Laboratory has been developed and is fully operational since 2019. To enrich its curriculum, a new practicum module has been developed, that can be immediately implemented and does not require any additional equipment or materials. To ensure safety in reactor kinetics and radiation protection, a safety analysis on the implementation of the practicum module has been conducted using MCNP and ORIGEN utilizing the current conditions of the reactor regarding its fuel burnup and control rod positions at a certain power level. Based on the results of the analysis, the practicum is safe to perform from a neutronic and radiation protection perspective. Given the long half-life and the large amount of radiation exposure that comes from activation products of iron, it is recommended that only cadmium, boron, graphite, and aluminum are allowed to be irradiated during the practicum.Keywords: Internet Reactor Laboratory, Activation Product, Radiation Protection, Reactor Safety

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Development of a personal-computer-based system for control rod worth determination in a nuclear reactor

IEEE Transactions on Nuclear Science ◽

10.1109/23.101247 ◽

1990 ◽

Vol 37 (6) ◽

pp. 2149-2155

Author(s):

S.A. Ansari ◽

B. Majid ◽

A. Hussain

Keyword(s):

Personal Computer ◽

Nuclear Reactor ◽

Control Rod ◽

Computer Based

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Finite Element Welding Residual Stress Analysis of CRDM Penetration Nozzles

Journal of Pressure Vessel Technology ◽

10.1115/1.4052451 ◽

2021 ◽

Vol 144 (1) ◽

Author(s):

Seung-Jae Kim ◽

Eui-Kyun Park ◽

Hong-Yeol Bae ◽

Ju-Hee Kim ◽

Nam-Su Huh ◽

...

Keyword(s):

Residual Stress ◽

Yield Strength ◽

Residual Stresses ◽

Nuclear Reactor ◽

Reactor Vessel ◽

Welding Residual Stress ◽

Alloy 600 ◽

Stress Distributions ◽

Control Rod ◽

Residual Stress Analysis

Abstract This article investigates numerically welding residual stress distributions of a tube with J-groove weld in control rod drive mechanisms of a pressurized nuclear reactor vessel. Parametric study is performed for the effect of the tube location, tube dimensions, and material's yield strength. It is found that residual stresses increase with increasing the inclination angle of the tube, and the up-hill side is the most critical. For thicker tube, residual stresses decrease. For material's yield strength, both axial and hoop residual stresses tend to increase with increasing the yield strength of Alloy 600. Furthermore, axial stresses tend to increase with increasing yield strength of Alloys 82/182.

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