1.1 本试验方法包括通过分析裂变反应的裂变产物（F.P.）来测量反应速率的程序 238 U（n，f）f.P。 1.2 该反应可用于测量能量约为1.5至7 MeV的中子，以及长达30至40年的辐照时间，前提是实际中描述的分析方法 E261 遵循。 1.3 实践中定义的等效裂变中子注量率 E261 可以确定。 1.4 实践中参考了其他快中子探测器的详细程序 E261 . 1.5 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。 1.6 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。 1.7 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 参考实践 E261 关于用裂变探测器测定快中子注量率的一般性讨论。 5.2 238 U可用作金属箔、金属丝或氧化物粉末（见指南 E844 ). 通常将其封装在适当的容器中，以防止材料的损失和污染 238 铀及其裂变产物。 5.3 可以分析一种或多种裂变产物。相关裂变产物的相关数据见 表1 和 表2 . （A） 括号中的光面数是最后一位数字中正负不确定性的大小。 （B） 具有 137m Ba（2.552分钟）处于平衡状态。 （C） 建议的半衰期和γ辐射概率取自参考文献 ( 3. ) 在设定推荐裂变产额时推荐的数据。 （D） 女儿的概率 140 洛杉矶衰变。 （E） 这是 140 洛杉矶/ 140 到达后Ba 瞬态平衡（t ≥ 19天）。 （A） 杰夫-3.1/3.1.1放射性衰变数据和裂变产额子库，杰夫报告20，经合组织2009，核能机构 ( 5. ) . （B） 所有产量数据以百分比形式给出；RC表示累积收益率；RI代表独立收益率。 5.3.1 137 Cs公司- 137m Ba经常用于长时间辐照。放射性产品 134 Cs和 136 Cs可能存在，这可能会干扰0.662 MeV的计数 137 Cs公司- 137m Ba伽马射线（见试验方法 E320 ). 5.3.2 140 文学学士- 140 对于短辐照，经常选择La（见试验方法 E393 ). 5.3.3 95 锆可以在化学分离后直接计数，也可以与其子体一起计数 95 Nb使用高分辨率伽马探测器系统。 5.3.4 144 Ce是一种高产量裂变产物，适用于2到3年的辐照。 5.4 有必要围绕 238 U监视器带有热中子吸收器，以最大限度地减少来自大量 235 U在 238 U目标和来自 239 Pu来自（n，γ）反应 238 U材料。测定 239 预计Pu浓度会有显著贡献。 5.4.1 轻水反应堆中子活化产物生产裂变产物 239 Pu被计算为无关紧要（<2 %), 相比之下 238 U（n，f），在快中子下辐照12年( E >1 MeV）1的注量率 × 10 11 厘米 −2. ·s −1. 前提是 238 U被热中子屏蔽（参见指南图2 E844 ). 5.4.2 光核反应（即，（γ，f）反应）产生的裂变产物，而在附近可忽略不计- 动力和研究堆芯，可用于深水渗透 ( 6. ) . 4. 5.5 中子注量测量值与 238 U裂变和 54 Fe（n，p） 54 锰反应已得到证实 ( 7. ) . 反应 238 U（n，f）f.P.是有用的，因为它比大多数阈值探测器对更大范围的中子能量作出响应。 5.6 这个 238 实践的表3给出了几个基准中子场中U裂变中子谱的平均截面 E261 . 指南中给出了最新推荐横截面的来源 E1018 . 在这种情况下 238 U（n，f）f.P.反应，建议的横截面来源是ENDF/B-VI释放8横截面（MAT = 9237) ( 8. ) . 图1 显示了快中子反应的推荐横截面与中子能量的关系图 238 U（n，f）f.P。 图1 ENDF/B-VI横截面与能量的关系 238 U（n，f）f.P.反应 注1: 数据取自评估的核数据文件ENDF/B- 六、 而不是后来的ENDF/B-VII。这符合指南 E1018 ，第6.1节，因为后面的ENDF/B-VII数据文件不包括协方差信息。存在一些协方差信息 238 U，但这仅适用于大于1 MeV的能量。有关更多详细信息，请参阅参考文件H节 9 .

1.1 This test method covers procedures for measuring reaction rates by assaying a fission product (F.P.) from the fission reaction 238 U(n,f)F.P. 1.2 The reaction is useful for measuring neutrons with energies from approximately 1.5 to 7 MeV and for irradiation times up to 30 to 40 years, provided that the analysis methods described in Practice E261 are followed. 1.3 Equivalent fission neutron fluence rates as defined in Practice E261 can be determined. 1.4 Detailed procedures for other fast-neutron detectors are referenced in Practice E261 . 1.5 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard. 1.6 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. 1.7 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 Refer to Practice E261 for a general discussion of the determination of fast-neutron fluence rate with fission detectors. 5.2 238 U is available as metal foil, wire, or oxide powder (see Guide E844 ). It is usually encapsulated in a suitable container to prevent loss of, and contamination by, the 238 U and its fission products. 5.3 One or more fission products can be assayed. Pertinent data for relevant fission products are given in Table 1 and Table 2 . (A) The lightface numbers in parentheses are the magnitude of plus or minus uncertainties in the last digit(s) listed. (B) With 137m Ba (2.552 min) in equilibrium. (C) The recommended half-life and gamma emission probabilities have been taken from the Reference ( 3 ) data that was recommended at the time that the recommended fission yields were set. (D) Probability of daughter 140 La decay. (E) This is the activity ratio of 140 La/ 140 Ba after reached transient equilibrium (t ≥ 19 days). (A) The JEFF-3.1/3.1.1 radioactive decay data and fission yields sub-libraries, JEFF Report 20, OECD 2009, Nuclear Energy Agency ( 5 ) . (B) All yield data given as a %; RC represents a cumulative yield; RI represents an independent yield. 5.3.1 137 Cs- 137m Ba is chosen frequently for long irradiations. Radioactive products 134 Cs and 136 Cs may be present, which can interfere with the counting of the 0.662 MeV 137 Cs- 137m Ba gamma rays (see Test Method E320 ). 5.3.2 140 Ba- 140 La is chosen frequently for short irradiations (see Test Method E393 ). 5.3.3 95 Zr can be counted directly, following chemical separation, or with its daughter 95 Nb using a high-resolution gamma detector system. 5.3.4 144 Ce is a high-yield fission product applicable to 2- to 3-year irradiations. 5.4 It is necessary to surround the 238 U monitor with a thermal neutron absorber to minimize fission product production from a quantity of 235 U in the 238 U target and from 239 Pu from (n,γ) reactions in the 238 U material. Assay of the 239 Pu concentration when a significant contribution is expected. 5.4.1 Fission product production in a light-water reactor by neutron activation product 239 Pu has been calculated to be insignificant (<2 %), compared to that from 238 U(n,f), for an irradiation period of 12 years at a fast-neutron ( E > 1 MeV) fluence rate of 1 × 10 11 cm −2 · s −1 provided the 238 U is shielded from thermal neutrons (see Fig. 2 of Guide E844 ). 5.4.2 Fission product production from photonuclear reactions, that is, (γ,f) reactions, while negligible near-power and research-reactor cores, can be large for deep-water penetrations ( 6 ) . 4 5.5 Good agreement between neutron fluence measured by 238 U fission and the 54 Fe(n,p) 54 Mn reaction has been demonstrated ( 7 ) . The reaction 238 U(n,f) F.P. is useful since it is responsive to a broader range of neutron energies than most threshold detectors. 5.6 The 238 U fission neutron spectrum-averaged cross section in several benchmark neutron fields is given in Table 3 of Practice E261 . Sources for the latest recommended cross sections are given in Guide E1018 . In the case of the 238 U(n,f)F.P. reaction, the recommended cross section source is the ENDF/B-VI release 8 cross section (MAT = 9237) ( 8 ) . Fig. 1 shows a plot of the recommended cross section versus neutron energy for the fast-neutron reaction 238 U(n,f)F.P. FIG. 1 ENDF/B-VI Cross Section Versus Energy for the 238 U(n,f)F.P. Reaction Note 1: The data is taken from the Evaluated Nuclear Data File, ENDF/B-VI, rather than the later ENDF/B-VII. This is in accordance with Guide E1018 , Section 6.1, since the later ENDF/B-VII data files do not include covariance information. Some covariance information exists for 238 U in the standard sublibrary, but this is only for energies greater than 1 MeV. For more details, see Section H of Ref 9 .