Active Standard ASTM E321 | Developed by Subcommittee: C26.05
Book of Standards Volume: 12.01
Historical (view previous versions of standard)
Significance and Use
The burnup of an irradiated nuclear fuel can be determined from the amount of a fission product formed during irradiation. Among the fission products, 148Nd has the following properties to recommend it as an ideal burnup indicator: (1) It is not volatile, does not migrate in solid fuels below their recrystallization temperature, and has no volatile precursors. (2) It is nonradioactive and requires no decay corrections. ( 3) It has a low destruction cross section and formation from adjacent mass chains can be corrected for. (4) It has good emission characteristics for mass analysis. (5) Its fission yield is nearly the same for 235U and 239Pu and is essentially independent of neutron energy (6). (6) It has a shielded isotope, 142Nd, which can be used for correcting natural Nd contamination. (7) It is not a normal constituent of unirradiated fuel.
The analysis of 148Nd in irradiated fuel does not depend on the availability of preirradiation sample data or irradiation history. Atom percent fission is directly proportional to the 148Nd-to-fuel ratio in irradiated fuel. However, the production of 148Nd from 147Nd by neutron capture will introduce a systematic error whose contribution must be corrected for. In power reactor fuels, this correction is relatively small. In test reactor irradiations where fluxes can be very high, this correction can be substantial (see Table 1).
The test method can be applied directly to U fuel containing less than 0.5 % initial Pu with 1 to 100 GW days/metric ton burnup. For fuel containing 5 to 50 % initial Pu, increase the Pu content by a factor of 10 to 100, respectively in both reagents 5.3 and 5.4.
TABLE 1 K Factors to Correct 148Nd for 147Nd Thermal Neutron CaptureA
|Total Neutron Flux,|
|Total Neutron Exposure, ϕI (neutrons/cm 2)|
|1 × 10 20||3 × 10 20||1 × 10 21||2 × 10 21||3 × 10 21|
|3 × 101 2||0.9985||0.9985||0.9985||0.9985||0.9985|
|1 × 101 3||0.9956||0.9952||0.9950||0.9950||0.9950|
|3 × 101 3||0.9906||0.9870||0.9856||0.9853||0.9852|
|1 × 101 4||0.9858||0.9716||0.9598||0.9569||0.9559|
|3 × 101 4||0.9835||0.9592||0.9187||0.9008||0.8941|
|1 × 101 5||0.9826||0.9526||0.8816||0.8284||0.8006|
A Assuming continuous reactor operation and a 274 ± 91 barn 1 47Nd effective neutron absorption cross section for a thermal neutron power reactor. This cross section was obtained by adjusting the 440 ± 150 barn 147Nd cross section (7) measured at 20°C to a Maxwellian spectrum at a neutron temperature of 300°C.
1.1 This test method covers the determination of stable fission product 148Nd in irradiated uranium (U) fuel (with initial plutonium (Pu) content from 0 to 50 %) as a measure of fuel burnup (1-3).
1.2 It is possible to obtain additional information about the uranium and plutonium concentrations and isotopic abundances on the same sample taken for burnup analysis. If this additional information is desired, it can be obtained by precisely measuring the spike and sample volumes and following the instructions in Test Method E267.
1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.
2. Referenced Documents (purchase separately) The documents listed below are referenced within the subject standard but are not provided as part of the standard.
D1193 Specification for Reagent Water
E180 Practice for Determining the Precision of ASTM Methods for Analysis and Testing of Industrial and Specialty Chemicals
E244 Test Method for Atom Percent Fission in Uranium and Plutonium Fuel (Mass Spectrometric Method)
E267 Test Method for Uranium and Plutonium Concentrations and Isotopic Abundances
ICS Number Code 27.120.30 (Fissile materials and nuclear fuel technology)
UNSPSC Code 15131500(Nuclear fuel)