JULY 1994 ECN-R--94-011 energy innovation ON THE SAFETY OF THE ALMR Some physics aspects A.J.JANSSEN CL 2 7 Si t 0 The Netherlands Energy Research Foundation ECN Het Energieonderzoek Centrum Nederland (ECN) is is the leading institute in the Netherlands for energy het centrale instituut voor onderzoek op energie- research. ECN carries out basic and applied research gebied in Nederland. ECN verricht fundamenteel en in the fields of nuclear energy, fossil fuels, renewable toegepast onderzoek op het gebied van kernenergie, energy sources, policy studies, environmental aspects fossiele-energiedragers, duurzame energie, beleids- of energy supply and the development and application studies, milieuaspecten van de energievoorziening en of new materials. de ontwikkeling en toepassing van nieuwe materialen. ECN employs more than 900 staff. Contracts are Bij ECN zijn ruim 900 medewerkers werkzaam. 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Box 1 Postbus l NL-1755ZG Petten 1755 ZG Petten the Netherlands Telefoon : (02246) 49 49 Telephone : +31 2246 49 49 Fax :(02246) 44 80 Fax : +31 2246 44 80 Dit rapport is te verkrijgen door het overmaken van This report is available on remittance of Dfl. 35 to: f35,-- op girorekening 3977703 ten name van: ECN, General Services, ECN, Algemene Diensten Petten, the Netherlands te Petten Postbank account No. 3977703. onder vermelding van het rapportnummer. Please quote the report number. © Netherlands Energy Research Foundation ECN © Energieonderzoek Centrum Nederland JULY 1994 ECN-R-94-011 *DE008128003* KSO01932830 R: FI DE008128003 ON THE SAFETY OF THE ALMR Some physics aspects A.J. JANSSEN ABSTRACT r j Reactivity coefficients of the Advanced Liquid Metal Reactor (ALMR) are defined and iterpreted. An analysis is made of potentially severe accidents in the ALMR. From this analysis conclusions are drawn with regard to the desired values of the reactivity coefficients for an inherently safe reactor design. A discussion is devoted to the question whether the proposed ALMR designs fulfil the desired criteria. This work was performed as part of the ECN research program ENGINE ECN Nuclear Energy acknowledges General Electric ALMR Core Engineering Division for providing additional information. ECN-R--94-011 CONTENTS 1. INTRODUCTION 5 2. REACTIVITY COEFFICIENTS 7 3. QUASI-STATIC ANALYSIS OF ACCIDENTS 13 3.1 Introduction 13 3.2 Influence of reactivity coefficients 13 3.3 Choices for an inherently safe design 15 3.4 Some calculation results 17 4. DISCUSSION 23 5. CONCLUSION 27 6. LITERATURE 29 ECN-R--94-011 ECN-R--94-011 1. INTRODUCTION The safety of modern nuclear power plants is based on principles like redundancy, diversity, defence-in-depth, etc. The extensive application of engineered safety features leads to higlily complex systems, entailing an increasing risk of human failure in the operation of these systems. In recent years new reactor designs have evolved which are characterized by simplification thanks to the application of inherent safety features. "Inherent safety" is defined as the elimination of inherent hazards through fundamental conceptual design choices, such that the reactor remains in a safe condition on the basis of laws of nature in all conceivable circumstances; no human interference, no triggering signals, and no supply of external energy are required to remain in a safe condition [22]. The term "passive safety" is also used frequently. Inherent safety refers to the self-control of the primary processes, whereas passive safety features come into operation (also without active triggering or energy supply) in case of a strong deviation of normal process behaviour. Since some remote failure mechanisms will always remain, no reactor concept can be qualified as completely inherently safe. However, specific safety features can be qualified as such, depending on the plant's design and response to accident initiators. In the Netherlands, the following three generations of reactor designs are distinguished by their (increasing) use of inherent safety features: 1. Evolutionary designs, based on existing reactors, which arc already now available (e.g. the Westinghouse APWR). 2. Advanced designs: more passive and less complex systems, e.g. the SBWR of General Electric. 3. Innovative designs, inherently safe to a high degree, which arc still under development. To this third generation belong, e.g., PIUS (which stands for Process Inherent Ultimate Safety) and PRISM (which stood for Power Reactor Inherently Safe Module). PRISM is an example of an "Advanced Liquid Metal Reactor" (ALMR). A judgement of the inherent safety of the ALMR may be based on its response to all credible accidents, assuming that all active safety provisions will fail. Many references can be found in the literature which address the safety of the ALMR. In almost all references three types of accidents are considered which - thanks to specific properties of the ALMR - develop without severe consequences, even without automatic or operator actions such as the insertion of control rods. These three accidents (which all belong to the group of Anticipated Transients Without Scram, ATWS) arc: 1. LOHS - Loss Of Heat Sink without scram. A pump failure in the water/steam circuit could be the cause of this accident. 2. LOF - Loss Of Flow without scram. A pump failure in the primary sodium circuit could be the cause of this accident. ECN-R--94-011 On the Safety of the ALMR 3. TOP - Transient Over Power without scram, caused by the unintended withdrawal of a control rod. Most papers which address these accidents in an ALMR are from General Electric (GE) and Argonnc National Laboratory (ANL). GE considers the PRISM design (about 450 MWth), ANL has considered a somewhat larger concept, about the size of SAFR (Safe Advanced Fast Reactor, ca 900 MWth), a concept of Rockwell International. These papers contain the results of transient calculations. The explanation and interpretation of these results are sometimes rather brief or even confusing: It is generally accepted that a negative Doppler effect is of utmust importance for the safety of a nuclear reactor of any type; nevertheless, GE states that several accidents develop benignly in PRISM thanks to "the small positive Doppler effect". To appreciate such remarks it is necessary to understand the basic physics properties of the ALMR and to make use of them in an evaluation of the accidents mentioned above. Fortunately, a few papers of ANL [12,14] consider such basic physics properties. We will develop our thoughts along the lines set out in these papers. In chapter 2 several reactor parameters are introduced which are connected to the safety of the ALMR. In chapter 3 we will use these parameters in an analysis of the accidents mentioned above. A comparison is made between two alternative designs - one with metallic fuel and one with oxide fuel. ECN-R-94-011 2. REACTIVITY COEFFICIENTS In reactor physics one usually defines reactivity coefficients which arc connected to changes of fundamental physics parameters. For the ALMR, a small sodium-cooled pool-type fast reactor, we can mention: 1. aD, the Doppler coefficient, which is connected to the fuel temperature. Doppler broadening of the nuclear resonances due to a temperature rise causes an increase of the neutron absorptions in the heavy nuclides, inducing a negative reactivity effect. 2. ctE, the axial fuel expansion coefficient. Axial expansion of the fuel rods reduces the volume-averaged fuel density in the core. This causes an increase of the neutron leakage from the core, especially in the radial direction. Axial fuel expansion is sometimes related directly to fuel temperature changes but in other cases it is dictated by coolant temperature changes, viz., if the fuel cannot expand freely within the cladding but remains stuck to the cladding of which the temperature is determined mainly by the coolant temperature. 3. aNa, the sodium density coefficient, which is connected to the coolant temperature. A reduction of the sodium density has two opposite effects. There will be less moderation of the neutrons; in the resulting harder neutron spectrum the ratio of neutron productions and neutron absorptions of the heavy nuclides is larger (this is the positive spectral effect). A decrease of the sodium density also makes the core more transparant for neutrons; the neutron leakage will therefore increase (this is the negative leakage effect). The net effect is positive in the core centre and negative at the outer boundaries; the volume-averaged effect is usually positive. (A third effect of a reduction of the sodium density, viz., a reduction of the neutron absorptions in the sodium, is of less importance.) It should be noted that in the accidents to be considered the coolant temperature increases will (hopefully) be so small that the sodium density will vary only slightly, entailing relatively small reactivity effects. If voids would occur (in case of sodium boiling or in case of other types of accidents) the (large) sodium void coefficient of the ALMR should be considered. 4. CCCR, the reactivity coefficient connected to the expansion of the "Control Rod Driveline". The control rods arc connected to the reactor closure head via the control rod drivelincs and are moved into the reactor core from above.