Значение и использование

5.1 Refer to Practice E261 for a general discussion of the determination of fast-neutron fluence rate with fission detectors.

5.2 ²³⁷Np is available as metal foil, wire, or oxide powder. For further information, see Guide E844. It is usually encapsulated in a suitable container to prevent loss of, and contamination by, the  ²³⁷Np and its fission products.⁴

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

TABLE 1 Recommended Nuclear Parameters for Certain Fission Products

Fission
Product

Parent
Half-Life^A (6)

Primary Radiation^A
(7) (keV)

γ Probability of
Decay^A (7)

Maximum
Useful
Irradiation
Duration

⁹⁵Zr

64.032 (6) d

724.192 (4)

0.4427 (22)

6 months

 

 

756.725 (12)

0.5438

 

⁹⁹Mo

65.94 (1) hr

739.500 (17)

0.1213 (22)

300 hours

 

 

777.921 (20)

0.0426 (8)

 

¹⁰³Ru

39.26 (2) d

497.085 (10)

0.910 (12)

4 months

¹³⁷Cs

30.05 (8) yr

661.657 (3)^B

0.8499 (20)^B

30–40 years

¹⁴⁰Ba–¹⁴⁰La

12.7527 (23) d

537.261 (3)

0.2439 (22)

1–1.5 months

 

 

1596.21 (4)

0.9540 (8)^C

 

 

 

 

1.1515^D

 

¹⁴⁴Ce

28.91 (5) d

133.515 (2)

0.1109 (19)

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 (1.67855 d) in transient equilibrium.TABLE 2 Recommended Fission Yields for Certain Fission Products^A

Fissile
Isotope

Neutron
Energy

Reaction
Product

Type
Yield

JEFF 3.1^B,A
Fission Yield (%)

²³⁷Np(n,f)

0.5 MeV

 ⁹⁵Zr

RC

5.6147 ± 2.7 %

 

 

 ⁹⁹Mo

RC

7.6118 ± 16.34 %

 

 

 ¹⁰³Ru

RC

5.4305 ± 12.7 %

 

 

 ¹³⁷Cs

RC

6.2654 ± 3.71 %

 

 

 ^137mBa

RI

1.4802e-3 ± 35.58 %

 

 

 ¹⁴⁰Ba

RC

5.9160 ± 3.82 %

 

 

 ¹⁴⁰La

RI

6.3568e-3 ± 36.68 %

 

 

 ¹⁴⁴Ce

RC

4.1230 ± 4.7 %

^A The JEFF-3.1/3.1.1 radioactive decay data and fission yields sub-libraries, JEFF Report 20, OECD 2009, Nuclear Energy Agency.^B All yield data given as a %; RC represents a cumulative yield; RI represents an independent yield.

5.3.1 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 E320).

5.3.2 ¹⁴⁰Ba-¹⁴⁰La is chosen frequently for short irradiations (see Test Method E393).

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

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

5.4 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.

5.4.1 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 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 (1).⁵

5.5 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.

5.6 The  ²³⁷Np fission neutron spectrum-averaged cross section in several benchmark neutron fields are given in Table 3 of Practice E261. Sources for the latest recommended cross sections are given in Guide E1018. 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.

FIG. 1 ENDF/B-VI Cross Section Versus Energy for the ²³⁷Np(n,f)F.P. ReactionNote 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 E1018 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) Область применения

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 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 and health practices and determine the applicability of regulatory limitations prior to use.