1.1 本试验方法涵盖稳定裂变产物的测定 148 辐照铀（U）燃料中的钕（初始钚（Pu）含量为0至50 %) 作为燃料燃耗的测量 ( 1- 3. ) . 2. 1.2 在进行燃耗分析的同一样品上，可以获得有关铀和钚浓度和同位素丰度的额外信息。如果需要此附加信息，可以通过精确测量峰值和样品体积并遵循测试方法中的说明来获得 E267 . 1.3 以国际单位制表示的数值应视为标准值。本标准不包括其他计量单位。 1.4 本标准并非旨在解决与其使用相关的所有安全问题（如有）。本标准的用户有责任在使用前制定适当的安全、健康和环境实践，并确定监管限制的适用性。 1.5 本国际标准是根据世界贸易组织技术性贸易壁垒（TBT）委员会发布的《关于制定国际标准、指南和建议的原则的决定》中确立的国际公认标准化原则制定的。 ====意义和用途====== 5.1 辐照核燃料的燃耗可以根据辐照期间形成的裂变产物的量来确定。在裂变产物中， 148 Nd具有以下特性，建议将其作为理想的燃耗指示器: 5.1.1 它不易挥发，在低于再结晶温度的固体燃料中不会迁移，并且没有挥发性前体。 5.1.2 它是非放射性的，不需要衰变校正。 5.1.3 它具有较低的破坏截面，相邻质量链的形成可以纠正。 5.1.4 它具有良好的质量分析排放特性。 5.1.5 它的裂变产额几乎与 235 U和 239 Pu和基本上不依赖于中子能量 ( 6. ) . 5.1.6 它有一个屏蔽同位素， 142 钕，可用于纠正自然钕污染。 5.1.7 它不是未辐照燃料的正常成分。 5.2 分析 148 辐照燃料中的钕不取决于辐照前样品数据或辐照历史的可用性。原子裂变百分比与 148 辐照燃料中的钕燃料比。然而，生产 148 Nd发件人 147 中子俘获钕将引入系统误差，必须对其贡献进行校正。在动力反应堆燃料中，这种修正相对较小。在通量可能非常高的试验反应堆辐照中，这种校正可能是实质性的（见 表1 ). （A） 假设反应堆连续运行和274±91仓 1. 4. 7. 热中子动力反应堆的有效中子吸收截面。该横截面是通过调整440±150谷仓获得的 1. 4. 7. Nd横截面 (7) 在20时测量 在300°C的中子温度下达到麦克斯韦谱 °C。 5.3 该试验方法可直接应用于含0.5%以下的铀燃料 % 初始功率因数为1至100 GW天/公吨燃耗。对于含5至50的燃油 % 初始Pu，将两种试剂中的Pu含量分别增加10到100倍 6.3 和 6.4 .

1.1 This test method covers the determination of stable fission product 148 Nd in irradiated uranium (U) fuel (with initial plutonium (Pu) content from 0 to 50 %) as a measure of fuel burnup ( 1- 3 ) . 2 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 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.4 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, health, and environmental practices and determine the applicability of regulatory limitations prior to use. 1.5 This international standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for the Development of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Committee. ====== Significance And Use ====== 5.1 The burnup of an irradiated nuclear fuel can be determined from the amount of a fission product formed during irradiation. Among the fission products, 148 Nd has the following properties to recommend it as an ideal burnup indicator: 5.1.1 It is not volatile, does not migrate in solid fuels below their recrystallization temperature, and has no volatile precursors. 5.1.2 It is nonradioactive and requires no decay corrections. 5.1.3 It has a low destruction cross section and formation from adjacent mass chains can be corrected for. 5.1.4 It has good emission characteristics for mass analysis. 5.1.5 Its fission yield is nearly the same for 235 U and 239 Pu and is essentially independent of neutron energy ( 6 ) . 5.1.6 It has a shielded isotope, 142 Nd, which can be used for correcting natural Nd contamination. 5.1.7 It is not a normal constituent of unirradiated fuel. 5.2 The analysis of 148 Nd in irradiated fuel does not depend on the availability of preirradiation sample data or irradiation history. Atom percent fission is directly proportional to the 148 Nd-to-fuel ratio in irradiated fuel. However, the production of 148 Nd from 147 Nd 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 ). (A) Assuming continuous reactor operation and a 274 ± 91 barn 1 4 7 Nd effective neutron absorption cross section for a thermal neutron power reactor. This cross section was obtained by adjusting the 440 ± 150 barn 1 4 7 Nd cross section (7) measured at 20 °C to a Maxwellian spectrum at a neutron temperature of 300 °C. 5.3 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 6.3 and 6.4 .