Staffan Jacobsson Svärd

Nondestructive Experimental Determination of the Pin-Power Distribution in Nuclear Fuel Assemblies

A need for validation of modern production codes with respect to the calculated pin-power distribution has been recognized. A nondestructive experimental method for such validation has been developed based on a tomographic technique. The gamma-ray flux distribution is recorded in each axial node of the fuel assembly separately, whereby the relative rod-by-rod content of the fission product 140Ba is determined. Measurements indicate that 1 to 2% accuracy (1σ) is achievable. A device has been constructed for in-pool measurements at reactor sites. The applicability has been demonstrated in measurements at the Swedish boiling water reactor (BWR) Forsmark 2 on irradiated fuel with a cooling time of 4 to 5 weeks. Data from the production code POLCA-7 have been compared to measured rod-by-rod contents of 140Ba. An agreement of 3.1% (1σ) has been demonstrated. It is estimated that measurements can be performed on a complete BWR assembly in 25 axial nodes within an 8-h work shift. As compared to the conventional method, involving gamma scanning of individual fuel rods, this method does not require the fuel to be disassembled nor does the fuel channel have to be removed. The cost per measured fuel rod is estimated to be an order of magnitude lower than the conventional method.

Verification and Determination of the Decay Heat in Spent PWR Fuel by Means of Gamma Scanning

Abstract Decay heat is an important design parameter at the future Swedish spent nuclear fuel repository. It will be calculated for each fuel assembly using dedicated depletion codes, based on the operator-declared irradiation history. However, experimental verification of the calculated decay heat is also anticipated. Such verification may be obtained by gamma scanning using the established correlation between the decay heat and the emitted gamma-ray intensity from 137Cs. In this procedure, the correctness of the operator-declared fuel parameters can be verified. Recent achievements of the gamma-scanning technique include the development of a dedicated spectroscopic data-acquisition system and the use of an advanced calorimeter for calibration. Using this system, the operator-declared burnup and cooling time of 31 pressurized water reactor fuel assemblies was verified experimentally to within 2.2% (1σ) and 1.9% (1σ), respectively. The measured decay heat agreed with calorimetric data within 2.3% (1σ), whereby the calculated decay heat was verified within 2.3% (1σ). The measuring time per fuel assembly was ˜15 min. In case reliable operator-declared data are not available, the gamma-scanning technique also provides a means to independently measure the decay heat. The results obtained in this procedure agreed with calorimetric data within 2.7% (1σ).

Method for Analyzing Fission Gas Release in Fuel Rods Based on Gamma-Ray Measurements of Short-Lived Fission Products

Fission gases are produced as a result of fission reactions in nuclear fuel. Most of these gases remain trapped within the fuel pellets, but some may be released to the fuel rod internal gas volume under certain conditions. This phenomenon of fission gas release is important for fuel performance since the released gases can degrade the thermal properties of the fuel rod fill gas and contribute to increasing fuel rod internal pressure. Various destructive and nondestructive methods are available for determining the amount of fission gas release; however, the current methods are primarily useful for determining the integrated fission gas release fraction, i.e., the amount of fission gas produced in the fuel that has been released to the free rod volume over the entire lifetime of a nuclear fuel rod. In this work, a method is proposed for determining the fission gas release that occurs during short irradiation sequences. The proposed method is based on spectroscopic measurements of gamma rays emitted in the decay of short-lived fission gas isotopes. Determining such sequence-specific fission gas release can be of interest when evaluating the fuel behavior for selected times during irradiation, such as during power ramps. The data obtained in this type of measurement may also be useful for investigating the mechanisms behind fission gas release for fuel at high burnup. The method is demonstrated based on the analysis of experimental gamma-ray spectra previously collected using equipment not dedicated for this purpose; however, the analysis indicates the feasibility of the method. Further evaluation of the method is planned, using dedicated equipment at the Halden Boiling Water Reactor.

Recent Progress in the Design of a Tomographic Device for Measurements of the Three-Dimensional Pin-Power Distribution in Irradiated Nuclear Fuel Assemblies

Abstract A tomographic technique for determination of the thermal power distribution in nuclear fuel assemblies is under development. The purpose is to provide an experimental validation tool for core simulation codes. Such codes are essential for the operation of nuclear power reactors, and validation is important in the process of improving and developing the codes as well as the fuel. The tomographic method is nonintrusive and offers large amounts of data within a normal revision shutdown. In earlier experimental investigations using a test platform, the method proved useful, demonstrating results of satisfying quality. However, the measuring setup also revealed nonfeasible properties related to transport, decontamination, and background radiation shielding. In this paper, the design of a new measuring device is presented. It is based on experiences from the test platform, but its size and weight make it advantageous regarding transports and decontamination. Moreover, the design inherently allows for more efficient background shielding. The latter has been investigated in a detailed study using the MCNP simulation code. The results confirm the high levels of background radiation observed in the test platform. It is also concluded that the shielding properties in the new design are sufficient.

Feasibility study of self powered neutron detectors in fast reactors for detecting local change in neutron flux distribution

Neutron flux monitoring systems form an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector systems have to be investigated with respect to practicality and feasibility according to the detection parameters. In this paper, we demonstrate the feasibility of using self powered neutron detectors as in-core detectors in fast reactors for detecting local changes in the neutron flux distribution. We show that the gamma contribution from fission products decay and activation of structural materials and sodium is very small compared to the fission gammas. Thus, it is possible for the in-core SPND signal to follow changes in local neutron flux as they are proportional to each other. This implies that the signal from an in-core SPND can provide dynamic information on the neutron flux perturbations occurring inside the reactor core.

Experimental study of background subtraction in Digital Cherenkov Viewing Device measurements

The Digital Cherenkov Viewing Device (DCVD) is an imaging tool used by authority inspectors for partial defect verification of nuclear fuel assemblies in wet storage, i.e. to verify that part of an ...

Partial Defect Evaluation Methodology for Nuclear Safeguards Inspections of Used Nuclear Fuel Using the Digital Cherenkov Viewing Device

Abstract This paper describes possible ways of analyzing and interpreting data obtained using the digital Cherenkov viewing device on spent nuclear fuel assemblies for the identification of partial defects in the fuel. According to the terminology of the International Atomic Energy Agency, partial defects refer to items, for instance, fuel assemblies, that are manipulated to the extent that a fraction of the fuel material is diverted or substituted. Analysis can be performed either by using a measure of the total light intensity or by identifying the light distribution pattern emanating from the spent nuclear fuel, the goal of either type of analysis being a quantitative measure that can be used in the data interpretation step. Two possible data interpretation alternatives are presented here: the threshold method and the hypothesis testing method. This paper summarizes some of the simulation studies and results that have been obtained, related to the two analysis and data interpretation methodologies.