D. Mohr

The experimental breeder reactor II inherent shutdown and heat removal tests — results and analysis

Abstract A test program is being conducted to demonstrate that a power-producing liquid-metal reactor (LMR) can (1) passively remove shutdown heat by natural convection, (2) passively reduce power in response to a loss of reactor flow, and (3) passively reduce power in response to a loss of the balance-of-plant heat sink. Measurements and pretest predictions confirm that natural convection is a reliable, predictable method of shutdown heat removal and suggest that safety-related pumps or pony motors are not necessary for safe shutdown heat removal in an LMR. Measurements from tests in which reactor flow and heat rejection to the balance of plant were perturbed show that reactivity feedbacks can passively control power and temperature. Data from these tests form a basis for additional tests including a complete loss of flow without scram and a complete loss of heat sink without scram.

Loss-of-primary-flow-without-scram tests: Pretest predictions and preliminary results

Abstract A series of tests in the Experimental Breeder Reactor No. 2 (EBR-II) has been concluded that investigated the effects of a complete loss of primary flow without scram. The development and preliminary study of these events is first discussed, including the test limits and controlling parameters. The results of two of the tests, SHRT 39 and 45, are examined in detail, although a compact summary of all the tests is included. The success in meeting the objectives of the test program served to verify that natural processes will shut down the reactor and maintain adequate cooling without control rod or operator intervention. The good comparison between predicted and measured results confirms that such events can be analyzed without elaborate codes if the basic processes are understood. Furthermore, recent studies suggest that the EBR-II results are characteristic of new innovative LMR designs being pursued in the U.S. that incorporate metallic driver fuel.

EBR-II unprotected loss-of-heat-sink predictions and preliminary test results☆

Abstract Two unprotected (i.e., no scram or plant protection system action) loss-of-heat-sink transients were performed on the Experimental Breeder Reactor-II in the Spring of 1986. One was initiated from full power (60 MW) and the other from half power. The loss of heat sink was accomplished in each test by essentially stopping the secondary-loop sodium coolant flow. Pretest predictions along with preliminary test results demonstrate that the reactor shuts itself down in a benign and predictable manner in which all of the reactor temperatures approach a quenching (or smothering) temperature at which the fission power goes to zero.

Safety analysis for the loss-of-flow and loss-of-heat sink without scram tests in EBR-II☆

Abstract This paper discusses the safety considerations and the analysis that was done to support the conduct of the Loss-of-Flow and Loss-of-Heat Sink Without Scram tests in EBR-II. The plant safety limits and the analysis codes and methods are presented. The impact of the tests and off-normal conditions on the fuel and plant structures is evaluated. Conclusions are that the test program had no impact on plant operations and the accumulated fuel damage was negligible.

Modification of EBR-II plant to conduct loss-of-flow-without-scram tests☆

Abstract This paper describes the details of and the philosophy behind changes made to the EBR-II plant in order to conduct loss-of-flow-without-scram tests. No changes were required to conduct loss-of-heat-sink-without-scram tests.

Studies of thermal-hydraulic phenomena in EBR-II

Abstract An experimental and theoretical program has been undertaken during the past several years with the objective of developing a well-documented understanding of steady-state and transient thermal-hydraulic behavior in EBR-II. The results of this effort have provided reactor designers and system modelers with needed integral-type demonstrations of important phenomena. This paper will discuss the particular problems of steady-state and transient hot channel peaking factors and plant operational characteristics impact upon natural circulation dynamics. Direct in-core experimental measurements have demonstrated that factors used for the prediction of peak coolant temperature rises at normal rated plant conditions may not be conservative due to pin-bundle distortions or inlet flow maldistributions, while those applied during loss-of-flow transients are most likely overconservative due to inter- and intrasubassembly phenomena. The importance of somewhat controllable parameters such as the sequence of primary and secondary pump trips and reactor scram, primary pump rundown times, and nominal operational power-to-flow ratio upon the dynamics of the transition from forced to natural convective flow are also presented.

Demonstration of EBR-II power maneuvers without control rod movement

Abstract A group of five plant inherent control tests was successfully conducted in November 1987 in the Experimental Breeder Reactor II. These tests demonstrated that the plant power of a metal-fueled reactor can be passively controlled over a large power range by slowly changing the primary flow and the reactor inlet temperature. These variables are, in turn, regulated by the primary pump speed, the secondary flow, and the turbine inlet pressure. In all tests, control rods were not used to regulate power. It was demonstrated that the plant power can be controlled with reasonable accuracy without using control rods when the reactivity feedback characteristics of the reactor are well understood and the plant controllers are adequately designed.

Results and Implications of the Experimental Breeder Reactor II Inherent Safety Demonstration Tests

Two milestone tests were conducted in the Experimental Breeder Reactor II (EBR-II), demonstrating some of the inherent safety features of a liquid-metal reactor. The first test was a loss of flow without scram and the second was a loss of heat sink without scram. Both tests were initiated from 100% power, and in both tests the reactor was shut down passively - by natural processes, principally thermal expansion - without automatic scram, operator intervention, or the help of special incore devices. The temperature transients during the tests were mild, as predicted, and there was no damage to the core or reactor plant structures. The test plus analysis demonstrated the feasibility of inherent passive shutdown for under cooling accidents and identified the more important features necessary for inherent shutdown and passive cooling. The results provide a technical basis for future experiments in EBR-II to investigate inherent safety for transient overpower accidents and to provide additional data for validation of computer codes used for design and safety analysis of inherently safe reactor plants.

The effect of primary pump coastdown characteristics on loss-of-flow transients without scram in EBR-II☆☆☆

Abstract A series of loss-of-flow (LOF) tests without scram (unprotected) is planned for the Experimental Breeder Reactor II (EBR-II) to demonstrate the inherent shutdown capability of the reactor during an LOF event. The purpose of this paper is to discuss in detail the unprotected LOF transient analysis, the validation of the EBR-II reactivity feedback modeling, and the significance of pump coastdown characteristics on peak reactor temperatures. The tests as designed are limited by the fuel-cladding eutectic temperature of the fuel elements, and in order to meet the required temperature limit, the initial power and flow of all the tests are 16.7 and 20% of their rated values, respectively. To further reduce peak temperature, the primary tank temperature is to be decreased to 338°C from the nominal 371°C. The results show that primary flow coastdown rate and the capacity of the auxiliary pump have dramatic effects on the reactor temperatures. The impact of secondary flow depends somewhat on test conditions. When the auxiliary pump is in operation, the effect of secondary flow behavior on the reactor temperature becomes less significant during an unprotected LOF event.

Predicted and measured response of the EBR-II plant to large steam pressure changes☆

Abstract Three tests were performed on the Experimental Breeder Reactor II (EBR-II) plant in which the steam pressure was ramped down by about 8, 16 and 32% of the initial 8806 kPa value, held constant, and then ramped back up to this value. Measured data from all three tests are provided along with a comparison with results from a numerical simulation of the down-ramp portion of the most severe test. The measured change in reactor inlet temperature was only about 25 to 30% of the change in steam drum saturation temperature. This relationship is very important in limiting power and temperature changes caused by steam system blowdowns in liquid metal reactor designs, such as the EBR-II, which utilize (large) negative temperature coefficients to enhance controllability and safety. These test results suggest that it appears possible to design an LMR plant in which reactivity feedbacks protect the reactor during a loss-of-heat-sink accident, without risking overly severe consequences during a steam pressure reduction accident.