Measurement system for calibrating the neutron apparatus of the control-rod system of transport nuclear power-generating systems

Atomic Energy ◽  
2000 ◽  
Vol 88 (5) ◽  
pp. 402-407
Author(s):  
V. D. Sevast'ynov ◽  
V. P. Yaryna ◽  
A. E. Volkov ◽  
V. D. Pavlov
Keyword(s):  
Measurement System ◽  
Nuclear Power ◽  
Control Rod ◽  
Rod System

Neutronic analysis of absorbing materials for the control rod system in reactor ALLEGRO

Kerntechnik ◽  
2016 ◽  
Vol 81 (4) ◽  
pp. 445-451
Author(s):  
F. Čajko ◽  
M. Sečanský ◽  
T. Chrebet ◽  
R. Zajac ◽  
P. Dařílek
Keyword(s):  
Control Rod ◽  
Rod System ◽  
Absorbing Materials

The reactor research and test facility

2020 ◽  
pp. 69-79
Author(s):  
Andrey S. KIRILLOV ◽  
Aleksandr P. PYSHKO ◽  
Andrey A. ROMANENKO ◽  
Valery I. YARYGIN
Keyword(s):  
Measurement System ◽  
Nuclear Power ◽  
Power Plants ◽  
Test Facility ◽  
Physical Parameters ◽  
Pumping System ◽  
Current State ◽  
History Of ◽  
Special Measurement ◽  
Space Nuclear Power

The paper describes an overview of the history of development and the current state of JSC “SSC RF-IPPE” reactor research and test facility designed for assembly, research and full-scale life energy tests of space nuclear power plants with a thermionic reactor. The leading specialists involved in development and operation of this facility are represented. The most significant technological interfaces and upgrade operations carried out in the recent years are discussed. The authors consider the use of an oil-free pumping system as part of this facility during degassing and life testing. Proposed are up-to-date engineering solutions for development of the automated special measurement system designed to record NPP performance, including volt-ampere characteristics together with thermophysical and nuclear physical parameters of a ground prototype of the space nuclear power plant. Key words: reactor research and test facility, thermionic reactor, life energy tests, oil-free pumping system, automated special measurement system, volt-ampere characteristics.

Comparative Analysis of the Control Rod Drive System

2012 ◽  
Vol 614-615 ◽  
pp. 1558-1561
Author(s):  
Wen Wei Han ◽  
Wei Shi Han ◽  
Qing Guo
Keyword(s):  
Comparative Analysis ◽  
Drive System ◽  
Hydraulic Control ◽  
Control Rod ◽  
Rod System ◽  
Drive Systems ◽  
Research Situation

This article has systematically summarized the recent research situation of control rod system in China and comparatively analyzed the features of a variety of control rod drive systems on a basis of brief introduction of common types of control rod drive system. It has been proposed to that the hydraulic control rod drive system have a great potential in a wide application concerning on ships, warships power reactors and protable desalination system.

RETRAN Application of Turbine Trip and Load Rejection of Startup Test Analysis for Lungmen ABWR

2009 ◽  
Author(s):  
Chen-Lin Li ◽  
Chiung-Wen Tsai ◽  
Chunkuan Shih ◽  
Jong-Rong Wang ◽  
Su-Chin Chung
Keyword(s):  
Control System ◽  
Water Level ◽  
Control Systems ◽  
Nuclear Power ◽  
Pressure Control ◽  
Reactor Power ◽  
Level Control ◽  
Bypass Flow ◽  
Control Rod ◽  
Recirculation Flow

This study used RETRAN program to analyze the turbine trip and load rejection transients of Taiwan Power Company Lungmen Nuclear Power Plant’s startup test at 100% power and 100% core flow operating condition. This model includes thermal flow control volumes and junctions, control systems, thermal hydraulic models, safety systems, and 1D kinetics model. In Lungmen RETRAN model, four steam lines are simulated as one line. There are four simulated control systems: pressure control system, water level control system, feedwater control system, and speed control system for reactor internal pumps. The turbine trip event, at above 40% power, triggers the fast open of the bypass valves. Upon the turbine trip, the turbine stop valves close. To minimize steam bypassed to the main condenser, recirculation flow is automatically runback and a SCRRI (selected control rod run in) is initiated to reduce the reactor power. The load rejection event causes the fast opening of the bypass valves. Steam bypass will sufficiently control the pressure, because of their 110% bypass capacity. A SCRRI and RIP runback are also initiated to reduce the reactor power. This study also investigated the sensitivity analysis of turbine bypass flow, runback rate of RIPS and SCRRI to observe how they affect fuel surface heat flux, neutron flux and water level, etc. The results show that turbine bypass flow has larger impacts on dome pressure than RIPS runback rate and SCRRI. This study also indicates that test criteria in turbine trip and load rejection transients are met and Lungmen RETRAN model is performing well and applicable for Lungmen startup test predictions and analyses.

ICONE15-10466 DESIGN OF THE CONTROL ROD SYSTEM FOR THE 2400 MW_ GENERATION IV GAS-COOLED FAST REACTOR

2007 ◽  
Vol 2007.15 (0) ◽  
pp. _ICONE1510-_ICONE1510
Author(s):  
G. Girardin ◽  
P. Coddington ◽  
F. Morin ◽  
G. Rimpault ◽  
R. Chawla
Keyword(s):  
Fast Reactor ◽  
Control Rod ◽  
Rod System

Dynamic Response Analysis of Control Rod Drive Mechanism Under the Stepping Impact Force for Nuclear Power Plants

2012 ◽  
Author(s):  
Xiaoyao Shen ◽  
Yongcheng Xie
Keyword(s):  
Dynamic Response ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Impact Force ◽  
Response Analysis ◽  
Control Rod ◽  
Drive Mechanism ◽  
Finite Element Software ◽  
The Impact

The control rod drive mechanism (CRDM) is an important safety-related component in the nuclear power plant (NPP). When CRDM steps upward or downward, the pressure-containing housing of CRDM is shocked axially by an impact force from the engagement of the magnetic pole and the armature. To ensure the structural integrity of the primary coolant loop and the functionality of CRDM, dynamic response of CRDM under the impact force should be studied. In this manuscript, the commercial finite element software ANSYS is chosen to analyze the nonlinear impact problem. A nonlinear model is setup in ANSYS, including main CRDM parts such as the control rod, poles and armatures, as well as nonlinear gaps. The transient analysis method is adopted to calculate CRDM dynamic response when it steps upward. The impact loads and displacements at typical CRDM locations are successfully obtained, which are essential for design and stress analysis of CRDM.

scholarly journals Ultrasonic Phased Array Evaluation of Control Rod Drive Mechanism (CRDM) Nozzle Interference Fit and Weld Region

2011 ◽  
Author(s):  
Anthony D. Cinson ◽  
Susan L. Crawford ◽  
Paul J. MacFarlan ◽  
Royce A. Mathews ◽  
Brady D. Hanson ◽  
...  
Keyword(s):  
Boric Acid ◽  
Nuclear Power ◽  
Phased Array ◽  
Crystal Form ◽  
Electric Discharge Machining ◽  
Interference Fit ◽  
Control Rod ◽  
The North ◽  
Ultrasonic Phased Array ◽  
Inner Diameter

Ultrasonic phased array data were collected on a removed-from-service CRDM nozzle specimen to assess a previously reported leak path. First a mock-up CRDM specimen was evaluated that contained two 0.076-mm (3.0-mil) interference fit regions formed from an actual Inconel CRDM tube and two 152.4-mm (6.0-in.) thick carbon steel blocks [1,2]. One interference fit region has a series of precision crafted electric discharge machining (EDM) notches at various lengths, widths, depths, and spatial separations for establishing probe sensitivity, resolution and calibration. The other interference fit has zones of boric acid (crystal form) spaced periodically between the tube and block to represent an actively leaking CRDM nozzle assembly in the field. Ultrasonic phased-array evaluations were conducted using an immersion 8-element annular 5.0-MHz probe from the tube inner diameter (ID). A variety of focal laws were employed to evaluate the interference fit regions and J-grove weld, where applicable. Responses from the mock-up specimen were evaluated to determine detection limits and characterization ability as well as contrast the ultrasonic response differences with the presence of boric acid in the fit region. Nozzle 63, from the North Anna Unit-2 nuclear power plant, was evaluated to assess leakage path(s) and was destructively dismantled to allow a visual verification of the leak path(s).

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