Notes on the Lumped Backward Master Equation for the Neutron Extinction/Survival Probability
- Los Alamos National Laboratory
The expected or mean neutron number (or density) provides an adequate characterization of the neutron population and its dynamical excursions in most neutronic applications, in particular power reactors. Fluctuations in the neutron number, originating from the inherent randomness of neutron interactions and fission neutron multiplicities, are relatively small and ignorable for operational purposes, although measurements of the variance and time correlations provide valuable diagnostic information on fundamental reactor physics parameters. However, it is well known that there exist situations of great interest and importance in which a strictly deterministic description, or even one supplemented with a knowledge of low order statistical averages (variance, correlation), provides an incomplete and very unsatisfactory description of the state of the neutron population. These situations are marked by persistent large fluctuations in the neutron number where the emergence of a deterministic phase is suppressed. Such situations are strongly stochastic and therefore unpredictable (i.e., the mean is not representative of the actual population), and can arise either by design or by accident. Examples where the stochastic behavior of neutron populations must be taken into account include: nuclear weapon single-point safety assessment; criticality excursions in spent fuel storage and in the handling of fissile solutions in fuel fabrication and reprocessing; approach to critical under suboptimal reactor start-up conditions; preinitiation in fast burst research reactors; and weak nuclear signatures in the passive detection of nuclear materials. What distinguishes strongly stochastic neutronic systems from strongly deterministic systems is that, in the former, neutron multiplication occurs in the presence of weak neutron sources, such as spontaneous fission and background (cosmic) radiation. Weak sources (in a sense that can be made quite precise) lead to well separated fission chains (a fission chain is defined as the initial source neutron and all its subsequent progeny) in which some chains are short lived while others propagate for unusually long times. Under these conditions, fission chains do not overlap strongly and this precludes the cancellation of neutron number fluctuations necessary for the mean to become established as the dominant measure of the neutron population. The fate of individual chains then plays a defining role in the evolution of the neutron population in strongly stochastic systems, and of particular interest and importance in supercritical systems is the extinction probability, defined as the probability that the neutron chain (initiating neutron and its progeny) will be extinguished at a particular time, or its complement, the time-dependent survival probability. The time-asymptotic limit of the latter, the probability of divergence, gives the probability that the neutron population will grow without bound, and is more commonly known as the probability of initiation or just POI. The ability to numerically compute these probabilities, with high accuracy and without overly restricting the underlying physics (e.g., fission neutron multiplicity, reactivity variation) is clearly essential in developing an understanding of the behavior of strongly stochastic systems.
- Research Organization:
- Los Alamos National Laboratory (LANL), Los Alamos, NM (United States)
- Sponsoring Organization:
- DOE/LANL
- DOE Contract Number:
- AC52-06NA25396
- OSTI ID:
- 1045392
- Report Number(s):
- LA-UR-12-22638; TRN: US1203659
- Country of Publication:
- United States
- Language:
- English
Similar Records
Some Considerations on Stochastic Neutron Populations
Monte Carlo Direct Simulation Method for Initiation Probability Problems
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEFENSE
22 GENERAL STUDIES OF NUCLEAR REACTORS
97 MATHEMATICAL METHODS AND COMPUTING
ACCURACY
CRITICALITY
FABRICATION
FISSION
FISSION NEUTRONS
FLUCTUATIONS
MULTIPLICITY
NEUTRON SOURCES
NEUTRONS
NUCLEAR WEAPONS
PHYSICS
POWER REACTORS
PROBABILITY
PROGENY
RANDOMNESS
REACTOR PHYSICS
REACTOR START-UP
REPROCESSING
RESEARCH REACTORS
SAFETY
SPENT FUEL STORAGE
SPONTANEOUS FISSION