scholarly journals EVALUATION OF BEARING LOADS AND CONTACT STRESSES IN THE PRIMARY COOLANT CHECK VALVE OF THE ADVANCED TEST REACTOR.

1970 ◽  
Author(s):  
N.R. Byers
Keyword(s):  
Check Valve ◽  
Contact Stresses ◽  
Primary Coolant ◽  
Bearing Loads ◽  
Test Reactor

Thermal Safety Margin Calculation of the MP-2 Experiment in the Advanced Test Reactor

2020 ◽  
Author(s):  
Grant L. Hawkes
Keyword(s):  
Finite Element ◽  
Flow Instability ◽  
National Laboratory ◽  
Thermal Analyses ◽  
Safety Margin ◽  
Heat Rate ◽  
Element Model ◽  
Nucleate Boiling ◽  
Primary Coolant ◽  
Test Reactor

Abstract The Mini-Plate 2 (MP-2) irradiation test is a fueled experiment designed for irradiation in multiple test locations in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). The experiment is a drop-in test where small aluminum-clad fuel plate samples (mini plates) are cooled directly by the ATR Primary Coolant System (PCS) water. The MP-2 fuel plate experiment will be irradiated in several different irradiation locations of the ATR. This fueled experiment contains aluminum-clad fuel mini plates consisting of monolithic U-Mo. Four different types of fuel plates with fuel meat thickness and cladding are part of the MP-2 test. A thermal analysis has been performed on the MP-2 experiment. A method for calculating Departure from Nucleate Boiling Ratio (DNBR) and Flow Instability Ratio (FIR) during a reactivity transient using the commercial finite element and heat transfer code ABAQUS is discussed. At the start of an ATR cycle the heat generation rate of the fueled experiment is high and the heat rate multiplier from the outer shim control cylinders is low, while the reverse is true at the end of the ATR cycle. Thermal analyses at 10-day increments during the cycle calculate the DNBR and FIR during a reactivity transient. This technique calculates DNBR for the fuel plate surfaces and FIR for all water components for each finite element surface and node at various times during the ATR cycle. Heat rates vary with time during the transient calculations that are provided by a detailed physics analysis. Oxide growth on the fuel plates is also incorporated. Results from the transient calculations are displayed with the ABAQUS post processor. By calculating these parameters at each location in the finite element model, conservatism is replaced with accuracy. This allows for a greater margin for the thermal hydraulic safety parameters.

Assessing Advanced Test Reactor Process States Using Acoustic Measurement Infrastructure

2017 ◽  
Author(s):  
Vivek Agarwal ◽  
James A. Smith
Keyword(s):  
Nuclear Reactor ◽  
Signal Transmission ◽  
National Laboratory ◽  
Reactor Core ◽  
Operating Conditions ◽  
Acoustic Measurement ◽  
Primary Coolant ◽  
Test Reactor ◽  
Operation Cycle ◽  
Measurement Infrastructure

The core of any nuclear reactor presents a particularly harsh environment for sensors and instrumentations. The reactor core also imposes challenging constraints on signal transmission from inside the reactor core to outside of the reactor vessel. In this paper, an acoustic measurement infrastructure installed at the Advanced Test Reactor (ATR), located at Idaho National Laboratory, is presented. The measurement infrastructure consists of ATR in-pile structural components, coolant, acoustic receivers, primary coolant pumps, a data-acquisition system, and signal processing algorithms. Intrinsic and cyclic acoustic signals generated by the operation of the primary coolant pumps are collected and processed. The characteristics of the intrinsic signal can indicate the process state of the ATR (such as reactor startup, reactor criticality, reactor attaining maximum power, and reactor shutdown) during operation (i.e., real-time measurement). This paper demonstrated different in acoustic signature of the ATR under different operating conditions. In particular, ATR acoustic baseline is captured during typical operation cycle and during power axial locator mechanism operation cycle. The difference in two acoustic baseline is significant and highlights salient difference that are critical in the design and development of acoustically telemetered sensors.

Investigation on Iodine Release Behavior During the Operation of High Temperature Engineering Test Reactor (HTTR)

2013 ◽  
Author(s):  
Shohei Ueta ◽  
Hiroyuki Inoi ◽  
Yoshitaka Mizutani ◽  
Hirofumi Ohashi ◽  
Jin Iwatsuki ◽  
...  
Keyword(s):  
High Temperature ◽  
Local Area ◽  
Release Behavior ◽  
Primary Coolant ◽  
Energy Agency ◽  
Xenon Isotopes ◽  
Test Reactor ◽  
Forced Cooling ◽  
Iodine Release ◽  
Testing Plan

Japan Atomic Energy Agency (JAEA) has planned to investigate on iodine release behavior from fuel through the testing operation of High Temperature Engineering Test Reactor (HTTR) in order to contribute to the reasonable estimation of the radiation exposure necessary for the realization of HTGR in the future. In this test, the fractional release of iodine will be measured and evaluated by measuring xenon isotopes, the daughter nuclides of iodine isotopes, in the primary coolant sampling under the loss-of-forced cooling (LOFC) test by which the primary coolant circulator is shut down and/or the manual scram test of HTTR. In parallel, the local area of primary coolant circuit where iodine is plated-out will be evaluated. This paper describes the testing plan and the preliminary analytical study on the release behavior of iodine and xenon isotopes through the operation of HTTR.

Thermal Analysis of the FSP-1 Experiment in the Advanced Test Reactor

2016 ◽  
Author(s):  
Grant L. Hawkes ◽  
Nicolas E. Woolstenhulme
Keyword(s):  
Thermal Analysis ◽  
Flow Instability ◽  
National Laboratory ◽  
Safety Evaluation ◽  
Thermal Safety ◽  
Primary Coolant ◽  
Coolant Pump ◽  
Safety Margins ◽  
Enriched Uranium ◽  
Test Reactor

The U.S. High Performance Research Reactor Conversions fuel development team is focused on developing and qualifying the uranium-molybdenum (U-Mo) alloy monolithic fuel to support conversion of domestic research reactors to low enriched uranium. Several previous irradiations have demonstrated the favorable behavior of the monolithic fuel. The Full Scale Plate 1 (FSP-1) fuel plate experiment will be irradiated in the northeast (NE) flux trap of the Advanced Test Reactor (ATR). This fueled experiment contains six aluminum-clad fuel plates consisting of monolithic U-Mo fuel meat. Three different types of fuel plates with matching pairs for a total of six plates were analyzed. These three types of plates are: full burn, intermediate power, and thick meat. A thermal analysis has been performed on the FSP-1 experiment to be irradiated in the ATR at the Idaho National Laboratory (INL). A thermal safety evaluation was performed to demonstrate that the FSP-1 irradiation experiment complies with the thermal-hydraulic safety requirements of the ATR Safety Analysis Report (SAR). The ATR SAR requires that minimum safety margins to critical heat flux and flow instability be met in the case of a loss of commercial power with primary coolant pump coast-down to emergency flow. The thermal safety evaluation was performed at 26 MW NE lobe power to encompass the expected range of operating power during a standard cycle. Additional safety evaluations of reactivity insertion events, loss of coolant event, and free convection cooling in the reactor and in the canal are used to determine the response of the experiment to these events and conditions. This paper reports and shows that each safety evaluation complies with each safety requirement of the ATR SAR.

Creep Properties of Base Metal and Welded Joint of Hastelloy XR Produced for High-Temperature Engineering Test Reactor in Simulated Primary-Coolant Helium

1999 ◽  
Vol 36 (12) ◽  
pp. 1160-1166 ◽  
Author(s):  
Yuji KURATA ◽  
Tatsuhiko TANABE ◽  
Isao MUTOH ◽  
Hirokazu TSUJI ◽  
Keijiro HIRAGA ◽  
...  
Keyword(s):  
High Temperature ◽  
Base Metal ◽  
Welded Joint ◽  
Primary Coolant ◽  
Creep Properties ◽  
Test Reactor

Experimental Investigation of 14C in the Primary Coolant of the 10 MW High Temperature Gas-Cooled Reactor

Radiocarbon ◽  
2019 ◽  
Vol 61 (03) ◽  
pp. 867-884 ◽  
Author(s):  
F Xie ◽  
W Peng ◽  
J Cao ◽  
X Feng ◽  
L Wei ◽  
...  
Keyword(s):  
High Temperature ◽  
Sampling System ◽  
Primary Coolant ◽  
Carbon 14 ◽  
The Core ◽  
Purification System ◽  
Very High Temperature Reactor ◽  
Test Reactor ◽  
Production Mechanisms ◽  
High Temperature Gas

ABSTRACTThe very high temperature reactor (VHTR) is a development of the high-temperature gas-cooled reactors (HTGRs) and one of the six proposed Generation IV reactor concept candidates. The 10 MW high temperature gas-cooled reactor (HTR-10) is the first pebble-bed gas-cooled test reactor in China. A sampling system for the measurement of carbon-14 (14C) was established in the helium purification system of the HTR-10 primary loop, which could sample 14C from the coolant at three locations. The results showed that activity concentration of 14C in the HTR-10 primary coolant was 1.2(1) × 102 Bq/m3 (STP). The production mechanisms, distribution characteristics, reduction routes, and release types of 14C in HTR-10 were analyzed and discussed. A theoretical model was built to calculate the amount of 14C in the core of HTR-10 and its concentration in the primary coolant. The activation reaction of 13C has been identified to be the dominant 14C source in the core, whereas in the primary coolant, it is the activation of 14N. These results can supplement important information for the source term analysis of 14C in HTR-10 and promote the study of 14C in HTGRs.

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