John I. Sackett

Implications of the EBR-II inherent safety demonstration test

Abstract Extensive thermal-hydraulics testing at EBR-II culminated in the Inherent Safety Demonstration Test on April 3, 1986. This work may well lead to fundamental changes in the approach to the design and licensing of liquid-metal-cooled reactor (LMR) power plants. The EBR-II test program has thus far demonstrated (1) passive removal of decay heat by natural circulation, (2) passive reactor shutdown for a loss of flow without scram, and (3) passive reactor shutdown for a loss of heat sink without scram. Supporting analyses indicate that these characteristics can be incorporated into larger commercial LMRs and be used as the basis for a totally new passive control strategy. Analyses and tests are now in progress to show that LMRs with these characteristics and the passive control strategy are also inherently safe for unprotected overpower accidents.

Evolution of thermal-hydraulics testing in EBR-II

Abstract A thermal-hydraulics testing and modeling program has been underway at the Experimental Breeder Reactor-II (EBR-II) for 12 years. This work culminated in two tests of historical importance to commercial nuclear power, a loss of flow without scram and a loss of heat sink without scram, both from 100% initial power. These tests showed that natural processes will shut EBR-II down and maintain cooling without automatic control rod action or operator intervention. Supporting analyses indicate that these results are characteristic of a range of sizes of liquid metal cooled reactors (LMRs), if these reactors use metal driver fuel. This type of fuel is being developed as part of the Integral Fast Reactor Program at Argonne National Laboratory. Work is now underway at EBR-II to exploit the inherent safety of metal-fueled LMRs with regard to development of improved plant control strategies.

Flow measurement in sodium and water using pulsed-neutron activation: Part 1, theory

A mathematical analysis and experimental averaging techniques are developed for flow measurements that use pulsed neutron activation. Experimental time-weighting procedures are evaluated for the two extremes of laminar and turbulent flow. In addition, criteria are developed that reflect expected error from an experimental procedure. Experimental data and confirmation of weighting techniques follow in a companion paper.

The future of nuclear energy

Nothing will be as important to the people of the world in the next century and beyond as energy to provide clean water and electricity. But how will this need be met with the increasingly recognized need to substantially reduce carbon dioxide emissions? Nuclear power provides a compelling option, but it must meet certain requirements in order to gain public and political support. In this paper, those requirements are examined and the imperative for continued research into advanced nuclear power technologies is discussed.

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.

Flow measurement in sodium and water using pulsed-neutron activation: Part 2, experiment

A measurement of water flow on the Experimental Breeder Reactor II (EBR-II) evaporator downcomers, a measurement of sodium flow on the EBR-II secondary sodium system, and a calibration experiment at the University of Utah Water Research Laboratory are three experiments using the pulsed neutron activation (PNA) technique to determine flow rate. The EBR-II data permit calculation of flow rates and comparisons with instrumentation and the Water Research Laboratory data permit investigation of different weighting schemes for determining the flow rates. The PNA technique is an accurate and convenient procedure that yields flow rates without accompanying system disturbance. Pipe size is not a factor except that corrections may be needed for asymmetry of larger pipes. Accuracy is adequate for most applications and indicates the PNA technique is most useful as a calibration device.

Severe Accidents and Containment Considerations

The previous chapter considered transient sequences that were terminated without damage by self protecting features of the reactor system design. In this chapter, we provide a mostly qualitative approach to the sequential steps typically followed to evaluate hypothetical core-disruptive-accidents (HCDAs) with significant core damage. In addition, design consideration for containment and accommodation of large sodium fires are presented.

General Safety Considerations

Fast reactors exhibit some unique characteristics related to safety in comparison to thermal reactors. At first glance, it might appear that achieving exceptional safety in a fast reactor might be more challenging than in a thermal reactor, considering that sodium-cooled fast reactors (SFR) have a higher core power density, the neutron lifetime is shorter, the effective delayed neutron fraction is less, the core is not arranged in its most reactive configuration, the sodium void effect is usually positive, and sodium interacts rather violently with air or water. On the other hand, the boiling point of sodium is sufficiently high that the reactor can be operated near atmospheric conditions (eliminating the massive pressures required for water-cooled systems), sodium has a very high heat capacity and thermal conductivity, the neuron mean free path is sufficiently long that spatial power shifts are negligible, and xenon poisoning is a non-issue. Furthermore, it has been demonstrated that passive safety features can be more easily incorporated than in the thermal reactor systems.

Steady-State Natural Circulation Performance of the Experimental Breeder Reactor II Primary Heat Transport Circuit

The Experimental Breeder Reactor II is a sodium-cooled fast breeder reactor and is designed to operate at a thermal power of 62.5 MW and an electrical generation rate of 20 MW. In a continuing program devoted to the understanding of the thermal, hydraulic, and neutronic behavior of this reactor under both normal and off-normal operating conditions, a series of steady-state natural convection tests have been conducted. Instrumentation utilized for the control and observation of the reactor behavior during these experiments included both the normal plant sensors as well as those located in-core within a special fueled subassembly. The results of these measurements have been compared to the predictions of an analytical model of the entire primary heat transport circuit and satisfactory agreement was obtained.