Spent fuel measurement

Standard high rate portable planar detector for the isotopic analysis of plutonium samples. High rate plutonium and spent fuel measurements. 

High sensitivity underwater radiation survey meter covering the range from 0.1 mr/hour to 30000 rad.hour. Spent fuel measurements. Also for unattended monitoring (if coupled to computer). 

The reactivity of the fuel loadings was monitored with three detectors one proportional counter and two fission chambers. For spent-fuel measurements, three fission chambers were used. Approach-lo- critical and pulse-neutron source data were recorded during loading and unloading of fuel tubes. A k-eff value of 0.94 was obtained with a loading of 54 tubes of unirradiated fuel, compared with a value of 0.95 obtained with 91 tubes of fuel irradiated to an average exposure of 3020 MWd/MT- The loadings were limited to a maximum k-eff value of 0.97 established as the test safety iimit. ... 

A PWR can operate steadily for periods of a year or two without refueling. Uranium-235 is consumed through neutron irradiation uranium-238 is converted into plutonium-239 and higher mass isotopes. The usual measure of fuel bumup is the specific thermal energy release. A typical figure for PWR fuel is 33,000 MWd/t. Spent fuel contains a variety of radionucHdes (50) ... 

The NRC also imposes special security requirements for spent fuel shipments and transport of highly enriched uranium or plutonium materials that can be used in the manufacture of nuclear weapons. These security measures include route evaluation, escort personnel and vehicles, communications capabiHties, and emergency plans. State governments are notified in advance of any planned shipment within their state of spent fuel, or any other radioactive materials requiring shipment in accident-proof. Type B containers. 

This project placed encapsulated spent fuel elements from an experimental AEG reactor into storage holes drilled into the floor of the mine located in a salt bed. Valuable experimental information was obtained about the interaction between the waste form and the salt in which the waste was emplaced. It was in fact this experiment, conducted in 1968, which revealed that inclusions of moisture, or brine, in the salt beds have a tendency to migrate up a thermal gradient towards a heat source placed in the salt. Quantities of brine were measured as migrating to the deposited waste canisters and the interaction of this brine with the canis-tered material was observed. 

One measure of the hazard associated with this waste is the water dilution volume (m3). The water dilution volume is the volume of water needed to dilute a radionuclide to its maximum permissible concentration in water. A plot of the water dilution volume (WDV) for spent fuel is shown in Figure 16.11. 

Though the activity of the spent fuel falls by more than a factor of 1000 in the first thousand years (Fig. 16.3), the WDV falls more slowly. This is due to the hazards posed by the long-lived a emitters in the spent fuel. For times greater than 500 years, the actinide radio toxicity prevails. On a time scale of 104-106 years, the WDV values approach those of the original ore used to make the reactor fuel. (There are limitations to this measure of hazard because of how radionuclides enter the biosphere and are concentrated.)... 

After fission of uranium 235, the radionuchdes produced in the spent fuel have cesium, strontium, iodine, and other radionuclides of very long half-lives that can be a danger. The other radio wastes include contaminated filters, wiping rags, solvents, protective clothes, hand tools, instruments and instrument parts, vials, needles, test tubes, and animal carcasses. Precautionary and preventive measures include ... 

The concentration of radioxenons in spent fuel is expected to be measurable. For example. 

Tracer techniques, for example, are used to obtain very small but representative and measurable samples of highly radioactive spent fuel solutions. One millilitre of the solution is then spiked with a known amount of uranium and plutonium tracer isotopes. A few microlitres of the spiked solution are dried and shipped to SAL. One to fifty nanograms of uranium or plutonium extracted from this tiny sample are sufficient for a complete analysis representing the composition of half a tonne of irradiated fuel with an accuracy of 0.3 to 0.5%. 

Separation of Actinides from the Samples of Irradiated Nuclear Fuels. For the purpose of chemical measurements of burnup and other parameters such as accumulation of transuranium nuclides in irradiated nuclear fuels, an ion-exchange method has been developed to separate systematically the transuranium elements and some fission products selected for burnup monitors (16) Anion exchange was used in hydrochloric acid media to separate the groups of uranium, of neptunium and plutonium, and of the transplutonium elements. Then, cation and anion exchange are combined and applied to each of those groups for further separation and purification. Uranium, neptunium, plutonium, americium and curium can be separated quantitatively and systematically from a spent fuel specimen, as well as cesium and neodymium fission products. 

The radioactivity of important radionuclides in spent fuel as a function of time after the fuel is removed from a reactor is indicated in Table 13.7. Figures 13.18 and 13.19 describe the water dilution volumes for the radionuclides in spent fuel and reprocessed waste as a function of time. (The water dilution volume is the volume of water needed to dilute the amount of a given isotope in the waste to a concentration safe for ordinary use.) Ordinant values in Figs. 13.18 and 13.19 can also be viewed as a measure of the radioactivities of the isotopes as a function of time. For both spent fuel and reprocessed waste, the chief source of radioactivity for the first 10 to 100 y is the fission products "Sr and " Cs. Thereafter, up to about 10,000 y. Am and (briefly) Pu isotopes are the dominant... 

Example 10-3. Cylindrical spent-fuel rods 0.5 in. in diameter and 7 ft long are removed after 3 yr from a reactor operating at a power level of 8,000 watts/lb fuel. After storage for 65 days, they are shipped in a lead-shielded, horizontally positioned cylindrical container, measuring 6.6 in. ID. The average density of the fuel including void volume is 7.84 g/om with a mass absorption coefficient tt/p = 0.045. The curie... 

The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit is to be taken for the reduced reactivity of burned or spent fuel relative to its original fresh composition, it is necessary to benchmark computational methods used in determining such reactivity worth against spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized- water reactors (PWR). 

Based on the data presented here and in the subsequent volumes of this report, the methodology employed in the generation of spent fuel isotopics and the criticality calculations using those isotopes is valid for bumup credit analyses. Further validation of the SCALE S AS2H sequence for generating spent fuel isotopics has been performed by comparison of calculated results with measured spent fuel chemical assay data. The criticality methods of CSAS/KENO V.a are validated against LWR-type fresh fuel critical experiments (both UO2 and MOX) in ref 19. 

As illustrated in Fig. 7, the MK-II spent fuel subassembly with a burnup of 62.5 GWd/t and cooling time of 5.2 years was measured. It was shown that the measured and MAGI calculated values were close except at the upper region of the fuel column. This difference was apparently due to the calculation error of neutron absorption by control rod. 

Chemical analysis of Nd was conducted at the hot cell facility. As Nd is one of the stable fission products and its fission yield is highly reliable, " Nd production obtained by destructive examination has been commonly used as a burnup index [6]. The calculated and measured burnup ratios for the MK-II spent fuel from 0.3 to 8.7 atom% is shown in Fig. 8. Measured results agreed with the MAGI burnup calculation within 5% in each row. 

The accuracy of decay heat calculations depends on the individual heat generation rate from fission product decay nuclides and actinides, and the burnup calculation for its production and transmutation. To obtain experimental data and to improve the accuracy of related calculations, the decay heat of MK-II spent fuel subassemblies was measured at the JOYO spent fuel storage pond [7], The fuel burnup was approximately 66 GWd/t and the cooling time was between 40 and 385 days. The measured decay heat is shown in Fig. 9. 

FIG. 9. Measured and calculated decay heat of JOYO MK-U spent fuel. 

All the fuel handling equipment of V-392 reactor plant (in-plant transport packing set for fresh and spent fuel, leak-tight bottles, bottles of defective assembly detection system) have the cells for fuel assemblies made of hexahedral tubes. This measure provides for improvement of nuclear safety under accident situations and also prevents mechanical damage of fuel assembly during its installation and withdrawal from the fuel handling equipment. 

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