WO2019239415A2

WO2019239415A2 - Method and apparatus for measuring nuclear fuel burnup

Info

Publication number: WO2019239415A2
Application number: PCT/IL2019/050670
Authority: WO
Inventors: Alexander Krakovich; Izhar NEDER; Shlomo Halfon
Original assignee: Soreq Nuclear Research Center
Priority date: 2018-06-14
Filing date: 2019-06-13
Publication date: 2019-12-19
Also published as: WO2019239415A3

Abstract

A method for determining an amount of a fissile element in a nuclear fuel assembly (FA), the method comprising: irradiating a given FA with a first beam of gamma radiation of known intensity and known spectrum; determining a first attenuated intensity equal to intensity of the gamma radiation from the first beam which passes through the given FA; and using the determined first attenuated intensity to determine an amount of the fissile element in the given FA and/or a bumup credit for the FA.

Description

METHOD AND APPARATUS FOR

MEASURING NUCUEAR FUEU BURNUP

REUATED APPUI CATIONS

[0001] The present application claims the benefit under 35 U.S.C. 119(e) of U.S. Provisional Applications 62/684,757 filed on June 14, 2018, the disclosure of which is incorporated herein by reference.

BACKGROUND

[0002] A nuclear power plant generates heat in a nuclear reactor to produce steam that is used to drive a turbine, which turns an electric generator to produce electricity. The heat is generated in a core of the reactor that may comprise a plurality of closely spaced fuel assemblies (FAs), each FA comprising nuclear fuel having an enriched concentration of fissile nuclei, such as uranium 235 (235pj) or plutonium 239 (2 9p_u^ housed in a closely packed array of fuel rods or fuel plates. The fuel rod/plate spacing, level of enrichment, amount of fuel in each FA and the number of FAs in the core, are determined so that they can be positioned sufficiently close to each other in the core to provide a mass and concentration of nuclear fuel that may be controlled to burn up and use fissile nuclei in the fuel in a critical nuclear fission chain reaction. Similarly, a research reactor which may be for example, a materials testing reactor (MTR) or a Training, Research, Isotopes, General Atomics (TRIGA) reactor uses fuel assemblies, comprising fuel rods or fuel plates, to provide a controllable critical nuclear fission chain reaction. However, research reactors are generally not used to provide electricity, and a portion of the neutrons and gamma rays produced by fission of nuclear fuel in the reactor is used for research applications rather than to support fission for producing electricity.

[0003] In a nuclear fission chain reaction, a fissile nucleus in the nuclear fuel undergoes a fission reaction in which it absorbs a neutron and splits into smaller nuclei (fission products) releasing large amounts of energy and new neutrons. At least one of the new neutrons interacts with and causes an additional fissile nucleus in the fuel to undergo fission in turn and repeat the cycle of releasing energy and inducing a further fission reaction to support the chain reaction. The chain reaction is self-sustaining and described as critical when each fission reaction generates, on average, one subsequent fission reaction, and a number of fission reactions and release of energy per second are substantially constant. The chain reaction extinguishes and is referred to as subcritical when each fission reaction generates on average less than one subsequent fission reaction. The chain reaction and release of energy grows exponentially and is referred to as supercritical when each fission reaction generates on average more than one subsequent fission reaction.

[0004] If the nuclear fuel in the core of a nuclear reactor goes subcritical and is not returned to criticality the reactor thermal power decreases rapidly, and the reactor shuts down. If the nuclear fuel goes supercritical and is not returned to criticality the release of energy may be so rapid and large as to extensively damage the reactor core. The reactor controls the nuclear fuel in the core to provide a critical chain reaction and prevent the fuel from going either subcritical or supercritical by dynamically controlling an extent to which control rods that absorb neutrons and prevent them from causing fission reactions are inserted into the core’s FAs.

[0005] When transporting and storing FAs or processing nuclear fuel, dynamic control of fuel criticality is not available, and care must be taken to assure that the nuclear fuel in the FAs does not go either critical or supercritical and release amounts of energy and radiation dangerous to life and the environment. Safety constraints mandated to provide safe handling of FAs are complex and expensive to implement, and typically require maintaining separation greater than a required minimum between FAs transported or stored together, and provision of suitable containment and shielding structures to prevent leakage of radiation from the FAs. The constraints become more stringent and more expensive to implement as the fissile enrichment of the nuclear fuel increases. Enrichment is highest and safety constraints most stringent for“fresh” FAs loaded with nuclear fuel that has not been burned in a reactor core. Enrichment decreases as the nuclear fuel in the FAs is burned and concentration of fissile nuclei in the fuel depleted as more and more of the fissile nuclei undergo fission.

[0006] To provide appropriate and cost-effective safety constraints for used FAs in which some of the fissile nuclei they contain has been burned or spent, and for which safety constraints may be relaxed, concentration of fissile nuclei they contain is determined to provide a measure referred to as a bumup credit. The burnup credit for an FA provides a measure of how much of the originally enriched fuel it contained when fresh has been spent and is used to configure the safety constraints that apply to handling the FA. Determining the burnup credit typically involves complex calculations and careful review of the use history of the FA. In the absence of burnup credit data, in order to ensure appropriate safety levels, it is assumed that no fissile nuclei have undergone fission. SUMMARY

[0007] An aspect of an embodiment of the disclosure relates to providing a method for determining depletion of nuclear fuel and burnup credit for an FA resulting from burnup of fissile material by measuring attenuation of gamma rays that have passed through the given FA. In an embodiment, to provide the determination, in addition to measuring attenuation for the given FA, attenuation of the gamma rays is measured for at least one reference FA having a known geometry and nuclear fuel composition. Optionally, the given FA and the at least one reference FA have a same geometry. Optionally, the at least one reference FA comprises a“fresh” FA loaded with fresh, unbumt nuclear fuel, and an“empty” FA that is empty of fuel. The attenuation measurements for the given FA and at least one reference FA are processed to determine fuel depletion and a bum-up credit for the given FA. In an embodiment, the energy of the gamma rays is chosen so that a difference between an attenuation coefficient of the nuclear fuel and fission products resulting from burning of the fuel is advantageously large.

[0008] This Summary is provided to introduce a selection of concepts in a simplified form that are further described below in the Detailed Description. This Summary is not intended to identify key features or essential features of the claimed subject matter, nor is it intended to be used to limit the scope of the claimed subject matter.

BRIEF DESCRIPTION OF FIGURES

[0009] Non-limiting examples of embodiments of the disclosure are described below with reference to figures attached hereto that are listed following this paragraph. Identical features that appear in more than one figure are generally labeled with a same label in all the figures in which they appear. A label labeling an icon representing a given feature of an embodiment of the disclosure in a figure may be used to reference the given feature. Dimensions of features shown in the figures are chosen for convenience and clarity of presentation and are not necessarily shown to scale.

[0010] Fig. 1 schematically shows apparatus operating to determine bumup credit for an FA, in accordance with an embodiment of the disclosure

[0011] Fig. 2 shows a line graph of attenuation coefficients as a function of gamma energy for different materials that may be present in a FA; [0012] Fig 3A, Fig 3B and Fig 3C schematically show apparatus acquiring gamma ray attenuation measurements for calibration FAs to determine burnup credit for an FA, in accordance with an embodiment of the disclosure; and

[0013] Fig. 4 schematically shows a configuration of an apparatus configured to determine burnup credit, in accordance with an embodiment of the disclosure.

DETAILED DESCRIPTION

[0014] In the description below, methods of determining burnup credit for an FA in accordance with an embodiment of the disclosure are described with reference to Fig. 1. Gamma ray energies that may be suitable for use in determining bumup credit in accordance with an embodiment are discussed with reference to Fig. 2, which shows a line graph of mass absorption coefficients for materials and nuclides commonly encountered in FAs and fission reactions in nuclear power plants and nuclear research reactors. A process for determining bumup credit using calibration FAs is illustrated in Figs. 3A-3C and discussed with reference to the figures. An example of an apparatus, hereinafter also referred to as a“BumUp meter”, for determining burnup credit in accordance with an embodiment of the disclosure is shown in and discussed with reference to Fig. 4.

[0015] In the discussion, unless otherwise stated, adjectives such as“substantially” and“about” modifying a condition or relationship characteristic of a feature or features of an embodiment of the disclosure, are understood to mean that the condition or characteristic is defined to within tolerances that are acceptable for operation of the embodiment for an application for which the embodiment is intended. Wherever a general term in the disclosure is illustrated by reference to an example instance or a list of example instances, the instance or instances referred to, are by way of non-limiting example instances of the general term, and the general term is not intended to be limited to the specific example instance or instances referred to. Unless otherwise indicated, the word“or” in the description and claims is considered to be the inclusive“or” rather than the exclusive or, and indicates at least one of, or any combination of more than one of items it conjoins.

[0016] Fig. 1 shows very schematically a BurnUp meter 20, acquiring a measure of gamma ray attenuation for a fuel assembly, FA 50, for use in determining bumup credit for the FA, in accordance with an embodiment of the disclosure. [0017] BumUp meter 20 optionally comprises a gamma ray source 22, a gamma ray detector 24, a collimator 26, and mechanical apparatus (not shown in Fig. 1) for positioning a portion of FA 50 between gamma ray source 22 and detector 24. In Fig. 1 FA 50 is schematically shown comprising fuel rods 52 for which nuclear fuel, represented by shading 54, in the fuel rods has by way of example been partially depleted by being used or“burnt up” in a nuclear reactor. Gamma ray source 22 provides a flux of gamma rays schematically represented by arrows 41 that are incident on FA 50. In the discussion, numeral 41, which refers to incident gamma rays, may also be used to refer to the flux of incident gamma rays.

[0018] Gamma rays 41 that enter FA 50 interact with material in the FA and are absorbed and scattered by the material. As a result, intensity of the incident gamma rays 41 attenuates with distance that the gamma rays propagate in the material of FA 50. Gamma rays 41 that survive to propagate all the way through FA 50, exit the FA in an attenuated exit flux of gamma rays, schematically represented by arrows 43. In an embodiment, collimator 26 collimates the attenuated exit flux of gamma rays 43 to shield detector 24 from extraneous, scattered gamma rays, and form a collimated exit beam of gamma rays 43 that is incident on detector 24. The detector provides a measure of intensity of the collimated beam and thereby of the flux of exit gamma rays 43. In the discussion, numeral 43, which refers to exit gamma rays, may also be used to refer to the flux of exit gamma rays.

[0019] In Fig. 1, a number of arrows representing attenuated exit flux 43 is less than a number of arrows representing incident flux 41 to schematically indicate that intensity of exiting flux 43 is attenuated with respect to that of incident gamma ray flux 41. However, whereas a greater number of arrows representing a gamma ray flux in Fig. 1 and Figs. 3A-3C discussed below indicates greater flux intensity, intensity of the flux may not be proportional to the number of arrows. As a result, a ratio between intensities of two gamma ray fluxes in the figures is not necessarily equal to a ratio between the numbers of arrows respectively representing the fluxes.

[0020] Attenuation of gamma rays in a material through which the gamma rays propagate is governed by exponential decrease as a function of a product of an amount of the material through which the gamma rays propagate times an attenuation coefficient. The attenuation coefficient is a function of the material and also of the energy of the gamma rays. If the amount of material is measured in units of mass the attenuation coefficient is referred to as a mass attenuation coefficient having units of area divided by mass. [0021] In an embodiment, energy of gamma rays 41 emitted by gamma ray source 22 is determined so that an attenuation coefficient of the gamma rays for the fissile nuclei of the nuclear fuel 54 in FA 50 is advantageously different from the attenuation coefficients for the fission products that result from fission of the fissile nuclei and from attenuation coefficients of structural material of the FA. As a result, attenuation of incident gamma rays 41 as a result of interaction with material in FA 50 is sensitive to changes in relative amounts of fissile nuclei and fissile products generated by fission of the fissile material in the FA. Intensity of attenuated gamma ray exit flux 43 that detector 22 measures relative to intensity of incident flux 41 is sensitive to and may be used to determine an amount of the fissile material in the FA. In particular, intensity measurements provided by detector 22 may be used to determine how much of an enriched concentration of fissile nuclei, originally present in FA when it was fresh, remains after the FA has been used and fuel in the FA burned. The measurements may therefore provide a measure of burnup credit for the FA.

[0022] In an embodiment, measurements of intensity of exit gamma flux 43 acquired by detector 22 may be used to determine attenuation of incident gamma ray flux 41. The determined attenuation may be compared to a calibration function that provides gamma ray attenuation as a function of concentration of fissile nuclei in fuel 54 to provide a burnup credit for FA 50. In an embodiment, BurnUp meter 20 may be operated to acquire attenuation measurements for at least one“calibration FA” similar to FA 50 and having a known concentration of fissile nuclei. BumUp meter 20 optionally uses the attenuation measurements for the at least one calibration FA and gamma ray attenuation measured for FA 50 to determine a bumup credit for FA 50.

[0023] By way of example, nuclear fuel in FA 50 is optionally U or Pu, which generate relatively abundant amounts of fission products such as Cesium 137 (137c_s) _anc[ Zirconium 90 (90zr). And, as discussed below, in an embodiment BurnUp meter 20 may be used to measure gamma ray attenuation for an FA, such as FA 20, in situ while the FA is immersed in water (H2O) in a reactor pool or a fuel storage facility. Mass attenuation coefficients as functions of gamma ray energy for the above-mentioned materials are respectively given by graph lines labeled U, Pu, Cs and Zr in line graph 200 shown in Fig. 2. For a range of energies between about 116 keV and about 400 keV indicated in line graph 200 by a dashed rectangle 201, the mass attenuation coefficient for U and Pu exhibits advantageous differences relative to the attenuation coefficients for Cs and Zr. For gamma ray energies in the indicated range, attenuation of the gamma rays is sensitive to changes in relative concentrations of U, Pu, Cs, Zr, and in an embodiment gamma ray source 22 is configured to provide gamma rays in the indicated energy range.

[0024] Any of various gamma emitting radioisotopes may be used to provide gamma rays in the 116-400 keV energy range. Gamma ray source 22 may for example comprise any one or any combination of more than one of Iridium 194 (194r-) Ytterbium 175 (175g¾) _and Lutetium 177

(1 7LU) which emit gamma rays at energies of 328 keV, 282 keV and 208 keV respectively. Each of the radioisotopes may be produced by irradiation of a progenitor nuclide with thermal neutrons and decays with a lifetime short enough so that practical concentrations of the radioisotope in gamma ray source 22 emit gamma rays having sufficient intensity so that detector 24 can acquire measurements of intensity of attenuated exit flux 43 that allow acceptably accurate determinations of burnup credit for FA 50. In an embodiment, the radioisotope may be irradiated with thermal neutrons that are available in a reactor core in which FA 50 may be located and can therefore be used in source 22 to enable BumUp meter 20 to make in-situ determinations of burnup credit for FA 50 while the FA is in the reactor core.

[0025] By way of example 776p_u has a very large cross section of about 2200 barns for neutron capture. Copious amounts of the radioisotope l^Fu _may |-,_c produced in source 22 by interaction of neutron flux available in a reactor core in which FA 50 is located with a concentration of 176_LU in the source. The half-lifetime of 177L_U is about 6.5 days, and as a result the produced 177L_U decays fast enough to provide a magnitude of incident flux 41 sufficient to provide practical determinations of burnup credit for FA 50.

[0026] Fig. 3A- Fig. 3C schematically illustrate BumUp meter 20 operating to determine bumup credit for FA 50 based on acquiring attenuation measurement for, optionally, two calibration FAs.

[0027] Fig. 3A is identical to Fig. 1 and schematically shows BurnUp meter 20 acquiring a measurement of intensity of attenuated gamma ray flux 43 exiting FA 50. Fig. 3B schematically shows BurnUp meter 20 acquiring measurement of attenuated gamma ray flux 63 for a first calibration FA 60 and a same intensity incident flux 41 as used to acquire the attenuation measurement for FA 50 shown in Fig. 3C. Calibration FA 60 is optionally substantially the same as FA 50 except that FA 60 is assumed to be fresh, and comprise as yet unused, that is unbumt, fuel 64 substantially the same as the original fuel in FA 50 prior to the fuel being burned. Fig. 3C schematically shows BumUp meter 20 acquiring a measurement of attenuated gamma ray flux 73 for a second calibration FA 70 that is substantially identical to FA 50 except that fuel rods 52 in FA 70 are empty.

[0028] As disclosed above and exemplified in figures 3A-C, calibration FAs such as fresh FAs and empty FAs may be used to calculate burnup. However, in accordance with an embodiment of the disclosure, calculation of bumup using calibration FAs is not limited to the use of two calibration FAs. Burnup for an FA 50 may be calculated using a fresh FA 60 as the calibration FA (with a correction due to minor actinide generation and fission). In an alternative embodiment of the disclosure, any FA for which an amount of nuclear fuel and the dimensions of the FA are known could be used as a calibration FA. Burnup in accordance with an embodiment can in general be determined based on any reference measurement of an FA having known composition and geometry. Reference may also be obtained by measuring the count rate from a reference FA directly immersed in water, as the composition and thickness of the attenuating water can be well defined.

[0029] By way of an example, for fuel assemblies for which the main fissile material is ²³⁵U (with negligible ²³⁸U or ²³⁹Pu fission during bumup), let the attenuation measurements for FA 50, calibration FA60 and calibration FA70 be represented by I5_Q, Ig_Q, and I7_Q, respectively. If bumup credit for FA 50 is represented by BUP50, then

BUP50 = c-ln(I₅₀/l60)/ln(l70/l60)· (Ό

In expression (1)“c” is a proportionality constant given by an expression,

c º [(p(af/a_a)(l-ppp/pu)]-l, (2) where h is the level of enrichment in fresh fuel, af and a_a are fission and attenuation cross sections respectively for 235pj mrr is a weighted average of the mass absorption coefficients of gamma rays for the fission products resulting from fission of 235pj and pp_j is the gamma ray mass absorption coefficient for 235pj. In an embodiment of the disclosure, for a case for which a mixture of fissile materials undergo fission during burn up, a modified equation is advantageously used. For example, for a mixture of different fissile nuclei equations 1 and 2 may be adjusted to account for differences in the cross sections of the different fissile nuclei and absorption coefficients of the different fission products.

[0030] Fig. 4 schematically shows a configuration of BumUp meter 20 for use in in-situ measurements of burnup credit for an FA such as FA 50, in accordance with an embodiment of the disclosure. BurnUp meter 20 includes a platform 413 and a rotating holder 402 for positioning the FA 50 in a horizontal state and providing suitable alignment of the FA 50 on platform 413. Gamma source 22 is housed in an, optionally aluminum, casing 411 which is mounted to a holder 410 fastened to an air- filled collimator 405. Air-filled collimator 405 may be between 2-3m long and have a diameter of less than lcm. In an embodiment of the disclosure, gamma source 22, FA 50, and collimator 405 are immersed in water of a water filled reactor pool 404 (water line indicated by line 416). Gamma source 22 may be activated to emit gamma rays while submerged in the reactor water by irradiation of a progenitor nuclide in the source with thermal neutrons in the core, Gamma radiation from source 22 that penetrates the FA travels through collimator 405 and is received by a gamma detector 24. on a sub-assembly. Gamma detector 24, and collimator 405 may be aligned to each other and mounted on a linear stage 414. Gamma detector 24 may be any suitable detector such as high purity Germanium detector or Cadmium Zinc Telluride (CdZnTe) detector. Stage 414 is optionally coupled to a rotating shaft 415 connected to a step motor 406 which rotates the shaft to position the stage at a desired location. Stability and co-ordination of disparate parts of the apparatus is achieved by various structural support components 403, 409 and 410.

[0031] In an embodiment of the disclosure, gamma radiation passing through an individual FA or a specific number of FAs may be measured. Instead of positioning the gamma source exterior to the FA, a gamma source 22 may be positioned inside the FA 50 itself such that there is just a single fuel rod/plate 52 or a specific number of fuel rods/plates between the gamma ray source 22 and the gamma ray detector 24. In an embodiment of the disclosure, the gamma source 22 may be specifically positioned in the FA by insertion into an empty control rod.

[0032] In an embodiment of the disclosure, the system enables advantageous design and utilization of transportation casks and waste repository storage spacing for both power and research reactor fuels.

[0033] In a further embodiment of the disclosure, the system also enables bumup measurement of FAs during nuclear fuel development or qualification.

[0034] There is therefore provided in accordance with an embodiment of the disclosure a method for determining an amount of a fissile element in a nuclear fuel assembly (FA), the method comprising: irradiating a given FA with a first beam of gamma radiation of known intensity and known spectrum; determining a first attenuated intensity equal to intensity of the gamma radiation from the first beam which passes through the given FA; and using the determined first attenuated intensity to determine an amount of the fissile element in the given FA and/or a bumup credit for the FA.

[0035] Optionally the method comprises irradiating a calibration FA having a known structure and comprising a known amount of the fissile element with a second beam of gamma radiation having a spectrum substantially the same as the spectrum of the first beam and a known intensity, and determining a second attenuated intensity equal to intensity of gamma radiation from the second beam that passes through the calibration FA.

[0036] The method may comprise using the determined second attenuated intensity to determine the amount of the fissile element in the given FA and/or the burnup credit. Optionally, the known amount of the fissile element in the calibration FA is equal to about zero. Alternatively, the known amount of fissile element in the calibration FA is equal to an amount of the fissile element that the calibration FA contains when loaded with fresh, unburnt nuclear fuel.

[0037] In an embodiment the structure of the calibration FA is substantially the same as the structure of the given FA.

[0038] In an embodiment using the second attenuated intensity comprises determining a ratio between the first and second attenuated intensities.

[0039] Optionally the method comprises: irradiating an additional calibration FA having a known structure and comprising fresh unbumt fuel comprising an amount of the fissile element that the given FA comprises when fully loaded with fresh unburnt fuel with a third beam of gamma radiation having a spectrum substantially the same as the spectrum of the first beam and a known intensity, and determining a third attenuated intensity equal to intensity of gamma radiation from the third beam that passes through the calibration; determining a first ratio between the second and third attenuated intensities; determining a second ratio between the first and third attenuated intensities; and using the first and second ratios to determine an amount of the using the first and second ratios to determine and amount of the fissile element in the given FA and/or the bumup credit. The method may comprise determining first and second logs of the ratios respectively. Optionally, the method comprises determining a ratio between the first and second logs.

[0040] In an embodiment, the gamma radiation in the first beam is characterized by an attenuation coefficient for the fissile element that is advantageously different from an attenuation coefficient for a fission product resulting from fission of the fissile element. Optionally, the gamma radiation is characterized by an attenuation coefficient for the fissile element that is advantageously different from an average of attenuation coefficients of the gamma radiation for fission products resulting from fission of the fissile element. The average may be a weighted average, weighted by relative quantities of the different fission products resulting from fission of the fissile element.

[0041] There is further provided in accordance with an embodiment of the disclosure apparatus for determining an amount of a fissile element in a nuclear fuel assembly (FA) and/or a bumup credit for the FA, the apparatus comprising: a source of gamma radiation of known intensity and known spectrum; a gamma radiation detector; a collimator mechanically coupled to the source and detector that collimates gamma radiation from the source into a beam that is incident on the detector; and a platform configured to hold a FA so that gamma radiation that the collimator collimates passes through the FA before reaching the detector.

[0042] In the description and claims of the present application, each of the verbs,“comprise” “include” and“have”, and conjugates thereof, are used to indicate that the object or objects of the verb are not necessarily a complete listing of components, elements or parts of the subject or subjects of the verb.

[0043] Descriptions of embodiments of the disclosure in the present application are provided by way of example and are not intended to limit the scope of the disclosure. The described embodiments comprise different features, not all of which are required in all embodiments. Some embodiments utilize only some of the features or possible combinations of the features. Variations of embodiments of the disclosure that are described, and embodiments comprising different combinations of features noted in the described embodiments, will occur to persons of the art. The scope of the invention is limited only by the claims.

Claims

1. A method for determining an amount of a fissile element in a nuclear fuel assembly (FA), the method comprising:

irradiating a given FA with a first beam of gamma radiation of known intensity and known spectrum;

determining a first attenuated intensity equal to intensity of the gamma radiation from the first beam which passes through the given FA; and

using the determined first attenuated intensity to determine an amount of the fissile element in the given FA and/or a burnup credit for the FA.

2. The method according to claim 1 and comprising irradiating a calibration FA having a known structure and comprising a known amount of the fissile element with a second beam of gamma radiation having a spectrum substantially the same as the spectrum of the first beam and a known intensity, and determining a second attenuated intensity equal to intensity of gamma radiation from the second beam that passes through the calibration FA.

3. The method according to claim 2 and comprising using the determined second attenuated intensity to determine the amount of the fissile element in the given FA and/or the bumup credit.

4. The method according to claim 3 wherein the known amount of the fissile element in the calibration FA is equal to about zero.

5. The method according to claim 3 wherein the known amount of fissile element in the calibration FA is equal to an amount of the fissile element that the calibration FA contains when loaded with fresh, unburnt nuclear fuel.

6. The method according to claim 3 wherein the structure of the calibration FA is substantially the same as the structure of the given FA.

7. The method according to claim 3 wherein using the second attenuated intensity comprises determining a ratio between the first and second attenuated intensities.

8. The method according to claim 4 and comprising:

irradiating an additional calibration FA having a known structure and comprising fresh unbumt fuel comprising an amount of the fissile element that the given FA comprises when fully loaded with fresh unburnt fuel with a third beam of gamma radiation having a spectrum substantially the same as the spectrum of the first beam and a known intensity, and determining a third attenuated intensity equal to intensity of gamma radiation from the third beam that passes through the calibration;

determining a first ratio between the second and third attenuated intensities; determining a second ratio between the first and third attenuated intensities; and using the first and second ratios to determine an amount of the using the first and second ratios to determine and amount of the fissile element in the given FA and/or the burnup credit.

9. The method according to claim 8 wherein using the first and second ratios comprises determining first and second logs of the ratios respectively.

10. The method according to claim 9 and comprising determining a ratio between the first and second logs.

11. The method according to claim 1 wherein the gamma radiation in the first beam is characterized by an attenuation coefficient for the fissile element that is advantageously different from an attenuation coefficient for a fission product resulting from fission of the fissile element.

12. The method according to claim 11 wherein the gamma radiation is characterized by an attenuation coefficient for the fissile element that is advantageously different from an average of attenuation coefficients of the gamma radiation for fission products resulting from fission of the fissile element.

13. The method according to claim 12 wherein the average is a weighted average, weighted by relative quantities of the different fission products resulting from fission of the fissile element.

14. Apparatus for determining an amount of a fissile element in a nuclear fuel assembly (FA) and/or a burnup credit for the FA, the apparatus comprising:

a source of gamma radiation of known intensity and known spectrum;

a gamma radiation detector;

a collimator mechanically coupled to the source and detector that collimates gamma radiation from the source into a beam that is incident on the detector; and

a platform configured to hold a FA so that gamma radiation that the collimator collimates passes through the FA before reaching the detector.