Neutron Multiplicity LAUR


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6. PASSIVE NEUTRON MULTIPLICITY COUNTING N. Ensslin, M. S. Krick, D. G. Langner, M. M. Pickrell, T. D. Reilly, and J. E. Stewart 6.1 INTRODUCTION 6.1.1 Purpose of the Chapter A nondestructive assay (NDA) technique for plutonium, called passive neutron multiplicity counting, has been developed as an extension of neutron coincidence counting (Ref. 1). The new technique has led to the design and fabrication of neutron multiplicity counters, one of which is pictured in Figure 6.1. The development of new neutron counters has been accompanied by advances in data-processing electronics, analysis algorithms, and data-analysis software. Development activities have been funded primarily by the Department of Energy (DOE) Office of Security Policy, Technology Development Branch. The new technology has led to significantly better measurement accuracy for plutonium metal, oxide, scrap, and residues. Fig. 6.1. Photo of the Plutonium Scrap Multiplicity Counter, used for accurate assays of plutonium metal, oxide, mixed oxide, or scrap . This chapter describes the principles of multiplicity counter design, electronics, and mathematics. Existing counters are surveyed, and their operating requirements and procedures are defined. Current applications to different plutonium material types are described and estimates of the expected assay precision and bias are given. 6.1.2 Definition of Neutron Multiplicity Counting Multiplicity is a word with a multiplicity of meanings! Our use of the word begins with the fact that an important NDA signature for plutonium is spontaneous fission, leading to the nearly simultaneous emission of multiple, indistinguishable neutrons. The number of neutrons emitted in spontaneous fission can vary from zero to eight. The distribution of the number of neutrons is LA-UR-07-1402 6-1
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called the multiplicity distribution. The multiplicity distributions for spontaneous fission in 240 Pu and the 2-MeV-neutron-induced fission for 239 Pu are illustrated in Figure 6.2. 012345678 0.0 0.1 0.2 0.3 0.4 Spontaneous Fission of 240 Pu 2MeV Induced Fission of 239 Pu Multiplicity Probability Fig. 6.2. The spontaneous fission multiplicity distribution for 240 Pu and the 2- MeV-neutron-induced fission multiplicity distribution for 239 Pu. Multiplicity counting sums up separately the number of 0, 1, 2, 3, 4, 5, 6, 7, etc. neutrons within the coincidence resolving time or “gate width” of the electronics package. This measures the multiplicity distribution of neutrons that are emitted, detected, and counted within the gate width. For this reason, the word multiplicity is specifically associated with the extension of conventional coincidence counting to the collection of higher-order multiples of neutrons. However, we also associate the word multiplicity with a special neutron counter design and with the mathematics of the data analysis process.
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