Determination of total plutonium content in spent nuclear fuel assemblies with the differential die-away self-interrogation instrument

Abstract

As a part of the Next Generation Safeguards Initiative Spent Fuel project, we simulate the response of the Differential Die-away Self-Interrogation (DDSI) instrument to determine total elemental plutonium content in an assayed spent nuclear fuel assembly (SFA). We apply recently developed concepts that relate total plutonium mass with SFA multiplication and passive neutron count rate. In this work, the multiplication of the SFA is determined from the die-away time in the early time domain of the Rossi-Alpha distributions measured directly by the DDSI instrument. We utilize MCNP to test the method against 44 pressurized water reactor SFAs from a simulated spent fuel library with a wide dynamic range of characteristic parameters such as initial enrichment, burnup, and cooling time. Under ideal conditions, discounting possible errors of a real world measurement, a root mean square agreement between true and determined total Pu mass of 2.1% is achieved.

Utilization of the Differential Die-Away Self-Interrogation Technique for Characterization and Verification of Spent Nuclear Fuel

“…However the review was held in the rather early stages of the investigation of these techniques and their capabilities have since been proven that were not believed to be possible during the review. For example, it was assumed that DDA and DDSI could not determine total elemental plutonium content, but that his since proven to be incorrect [33,34]. and SINRD which were expected to be less expensive.…”

“…Assuming that efficiency is absorbed into the calibration parameters, the eq. (11), i.e., the expression for determination of m pu , can be rewritten using only quantities directly measurable by the DDSI instrument as follows [34]:…”

“…It has been shown, by means of MCNP simulations, that total Pu content can be determined by DDA-and DDSI-based instruments [33,34] based on the known irradiation history of a given SFA, the total Pu content can be estimated through various burnup codes with an accuracy of 5%-10%. Such a large uncertainty may seem to be rather limiting; however, it should be noted that the NDA instrument used for Pu determination will likely use, for its own calibration, a calculated Pu content from a subset of the SFAs to be assayed.…”

“…When assaying induced spent fuel, the neutron source is self-contained in the abundant spontaneous fission isotopes such as 244 Cm and 240 Pu. The characteristic die-away time observed from different assemblies with varying initial enrichment, burnup, and cooling time allows us to determine SFA properties such as multiplication and total Pu content[9,34]. When assaying fresh fuel, a spontaneous fission driver source must be added to provide the interrogating neutrons since 235 U and 238 U spontaneous fission rates are too low for any practical use with the DDSI method.The DDSI instrument contains 563 He detectors in four stainless steel enclosures surrounding a central cavity.Figure 6-1 shows one of the open detector enclosures.…”

Data Mining Techniques to Estimate Plutonium, Initial Enrichment, Burnup, and Cooling Time in Spent Fuel Assemblies

“…1 predictor, total count rate during delayed neutron time window DG 4 predictors, 4 derived quantities estimating contribution of 239 Pu, 241 Pu, 235 U, and 238 U DDSI 2 to 10 predictors, magnitudes of fast and slow components of the Rossi-alpha distribution, die-away constants of exponential fits to those components, fast and slow area integrals under the Rossi-alpha distribution in the early (5-50) and late (100-200) time domains respectively, the correlated multiplication in the assembly from the early die-away time, the total number of neutrons detected per second (singles), and the slow area divided by the singles rate [Kaplan, 2014]. However, only early, fast, and slow die away times plus slow area influence the results significantly as shown in Table 5.…”