ASTM-E705 2008(Redline)

Refer to Practice E 261 for a general discussion of the determination of fast-neutron fluence rate with fission detectors.

²³⁷ Np is available as metal foil, wire, or oxide powder. For further information, see Guide E 844. It is usually encapsulated in a suitable container to prevent loss of, and contamination by, the ²³⁷ Np and its fission products.

One or more fission products can be assayed. Pertinent data for relevant fission products are given in Table 1 and Table 2.

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 ray (see Test Methods E 320).

¹⁴⁰ Ba- ¹⁴⁰ La is chosen frequently for short irradiations (see Test Method E 393).

⁹⁵ Zr can be counted directly, following chemical separation, or with its daughter ⁹⁵ Nb, using a high-resolution gamma detector system.

¹⁴⁴ Ce is a high-yield fission product applicable to 2- to 3-year irradiations.

It is necessary to surround the ²³⁷ Np monitor with a thermal neutron absorber to minimize fission product production from trace quantities of fissionable nuclides in the ²³⁷ Np target and from ²³⁸ Np and ²³⁸ Pu from (n, γ ) reactions in the ²³⁷ Np material. Assay of ²³⁸ Pu and ²³⁹ Pu concentration is recommended when a significant contribution is expected.

Fission product production in a light-water reactor by neutron activation products ²³⁸ Np and ²³⁸ Pu has been calculated to be insignificant (1.2 %), compared to that from ²³⁷ Np(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 ²³⁷ Np is shielded from thermal neutrons (see Fig. 2 of Guide E 844).

Fission product production from photonuclear reactions, that is, ( γ ,f) reactions, while negligible near-power and researchreactor cores, can be large for deep-water penetrations (1).

Good agreement between neutron fluence measured by ²³⁷ Np fission and the ⁵⁴ Fe(n,p) ⁵⁴ Mn reaction has been demonstrated (2). The reaction ²³⁷ Np(n,f) F.P. is useful since it is responsive to a broader range of neutron energies than most threshold detectors.

The ²³⁷ Np fission neutron spectrum-averaged cross section in several benchmark neutron fields are given in Table 3 of Practice E 261. Sources for the latest recommended cross sections are given in Guide E 1018. In the case of the ²³⁷ Np(n,f)F.P. reaction, the recommended cross section source is the ENDF/B-VI cross section (MAT = 9346) revision 1 (3) . Fig. 1 shows a plot of the recommended cross section versus neutron energy for the fast-neutron reaction ²³⁷ Np(n,f)F.P.

Note 1—The data are taken from the Evaluated Nuclear Data file, ENDF/B-VI, rather than the later ENDF/B-VII. This is in accordance with Guide E 1018 Guide for Application of ASTM Evaluated Cross Section Data File, 6.1. since the later ENDF/B-VII data files do not include covariance information. For more details see Section H of (10)

TABLE 1 Recommended Nuclear Parameters for Certain Fission Products

^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 Probability of daughter ¹⁴⁰ La decay.

^D With ¹⁴⁰ La (1.6781 d) in transient equilibrium.

TABLE 2 Recommended Fission Yields for Certain Fission Products ^A

^A Special issue on Evaluated Nuclear Data File ENDF/B-VII.0. ” Nuclear Data Sheets, J.K. Tull Editor. Vol. 107 December 2006. Data available on the ENDF/B-VII website at URL:http://www.nndc.bnl.gov/exfor/endf00.htm.

^B All yield data given as a %; RC represents a cumulative yield; RI represents an independent yield.

FIG. 1 ENDF/B-VI Cross Section Versus Energy for the ²³⁷ Np(n,f)F.P. Reaction

1. Scope

1.1 This test method covers procedures for measuring reaction rates by assaying a fission product (F.P.) from the fission reaction ²³⁷ Np(n,f)F.P.

1.2 The reaction is useful for measuring neutrons with energies from approximately 0.7 to 6 MeV and for irradiation times up to 30 to 40 years.

1.3 Equivalent fission neutron fluence rates as defined in Practice E 261 can be determined.

1.4 Detailed procedures for other fast-neutron detectors are referenced in Practice E 261.

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 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.

ASTM Standards

E170 Terminology Relating to Radiation Measurements and Dosimetry E181 Test Methods for Detector Calibration and Analysis of Radionuclides E261 Practice for Determining Neutron Fluence, Fluence Rate, and Spectra by Radioactivation Techniques E262 Test Method for Determining Thermal Neutron Reaction Rates and Thermal Neutron Fluence Rates by Radioactivation Techniques E320 Test Method for Cesium-137 in Nuclear Fuel Solutions by Radiochemical Analysis E393 Test Method for Measuring Reaction Rates by Analysis of Barium-140 From Fission Dosimeters E704 Test Method for Measuring Reaction Rates by Radioactivation of Uranium-238 E844 Guide for Sensor Set Design and Irradiation for Reactor Surveillance, E 706 (IIC) E944 Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, E 706 (IIA) E1005 Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance, E 706 (IIIA) E1018 Guide for Application of ASTM Evaluated Cross Section Data File, Matrix E706 (IIB)

Keywords

fission dosimeter; fission produc