Significance and Use
Refer to Practice E 261 for a general discussion of the determination of fast-neutron fluence rate with fission detectors.
237Np 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 237Np 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.
137Cs-137mBa is chosen frequently for long irradiations. Radioactive products 134Cs and 136Cs may be present, which can interfere with the counting of the 0.662 MeV 137Cs-137mBa gamma ray (see Test Methods E 320).
140Ba-140La is chosen frequently for short irradiations (see Test Method E 393).
95Zr can be counted directly, following chemical separation, or with its daughter 95Nb, using a high-resolution gamma detector system.
144Ce is a high-yield fission product applicable to 2- to 3-year irradiations.
It is necessary to surround the 237Np monitor with a thermal neutron absorber to minimize fission product production from trace quantities of fissionable nuclides in the 237Np target and from 238Np and 238Pu from (n,γ) reactions in the 237Np material. Assay of 238Pu and 239Pu concentration is recommended when a significant contribution is expected.
Fission product production in a light-water reactor by neutron activation products 238Np and 238Pu has been calculated to be insignificant (1.2 %), compared to that from 237Np(n,f), for an irradiation period of 12 years at a fast neutron (E > 1 MeV) fluence rate of 1 × 1011 cm−2·s−1, provided the 237Np 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 237Np fission and the 54Fe(n,p)54Mn reaction has been demonstrated (2). The reaction 237Np(n,f) F.P. is useful since it is responsive to a broader range of neutron energies than most threshold detectors.
The 237Np 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 237Np(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 237Np(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
|γ Probability of|
|95Zr||64.032 (6) d||724.192 (4)||0.4427 (22)||6 months|
|99Mo||2.7489 (6) d||739.500 (17)||0.1213 (22)||300 hours|
|777.921 (20)||0.0426 (8)|
|103Ru||39.26 (2) d||497.084 (6)||0.910 (12)||4 months|
|137Cs||30.3 (5) yr||661.657 (3)B||0.8510B||30–40 years|
|140Ba–140La||12.752 (5) d||537.261 (9)||0.2439 (23)||1–1.5 months|
|1596.21 (4)||0.954 (14)C|
|144Ce||289.91 (5) d||133.515 (2)||0.1109 (10)||2–3 years|
A The lightface numbers in parentheses are the magnitude of plus or minus uncertainties in the last digit(s) listed.
B With 137mBa (2.552 min) in equilibrium.
C Probability of daughter 140La decay.
D With 140La (1.6781 d) in transient equilibrium.
TABLE 2 Recommended Fission Yields for Certain Fission ProductsA
Fission Yield (%)
|237Np(n,f)||0.5 MeV|| 95Zr||RC||5.66915 ± 2 %|
| 99Mo||RC||6.11804 ± 4 %|
| 103Ru||RC||5.5583 ± 2.8 %|
| 137Cs||RC||6.25127 ± 2 %|
| 137mBa||RI||1.141e-3 ± 64 %|
| 140Ba||RC||5.48848 ± 2 %|
| 140La||RI||5.121e-3 ± 64 %|
| 144Ce||RC||4.13935 ± 2 %|
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 237Np(n,f)F.P. Reaction
1.1 This test method covers procedures for measuring reaction rates by assaying a fission product (F.P.) from the fission reaction 237Np(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.
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)
fission dosimeter; fission product; fission reaction rates; Neptunium-237; Fast neutron flux/fluence; Neptunium-237; Neutron activation reactions; Radioactivation--fast neutron flux; Threshold detectors--0.7 to 6 MeV;
ICS Number Code 17.240 (Radiation measurements); 27.120.30 (Fissile materials and nuclear fuel technology)
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