GB2600505A - Estimation method and monitoring system for plutonium concentration in uranium plutonium solution system based on neutron coincidence counting - Google Patents
Estimation method and monitoring system for plutonium concentration in uranium plutonium solution system based on neutron coincidence counting Download PDFInfo
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- WJWSFWHDKPKKES-UHFFFAOYSA-N plutonium uranium Chemical compound [U].[Pu] WJWSFWHDKPKKES-UHFFFAOYSA-N 0.000 title claims abstract description 86
- 229910052778 Plutonium Inorganic materials 0.000 title claims abstract description 55
- OYEHPCDNVJXUIW-UHFFFAOYSA-N plutonium atom Chemical compound [Pu] OYEHPCDNVJXUIW-UHFFFAOYSA-N 0.000 title claims abstract description 54
- 238000000034 method Methods 0.000 title claims abstract description 43
- 238000012544 monitoring process Methods 0.000 title claims abstract description 32
- 230000004992 fission Effects 0.000 claims abstract description 157
- 238000001514 detection method Methods 0.000 claims abstract description 130
- 230000002269 spontaneous effect Effects 0.000 claims abstract description 103
- 238000012937 correction Methods 0.000 claims abstract description 40
- 238000001228 spectrum Methods 0.000 claims abstract description 31
- 238000005259 measurement Methods 0.000 claims abstract description 16
- 238000010521 absorption reaction Methods 0.000 claims abstract description 10
- 238000007405 data analysis Methods 0.000 claims description 24
- OYEHPCDNVJXUIW-AHCXROLUSA-N Plutonium-240 Chemical compound [240Pu] OYEHPCDNVJXUIW-AHCXROLUSA-N 0.000 claims description 22
- 230000009257 reactivity Effects 0.000 claims description 16
- 238000013480 data collection Methods 0.000 claims description 12
- OYEHPCDNVJXUIW-VENIDDJXSA-N plutonium-238 Chemical compound [238Pu] OYEHPCDNVJXUIW-VENIDDJXSA-N 0.000 claims description 7
- 238000004364 calculation method Methods 0.000 claims description 6
- 238000011065 in-situ storage Methods 0.000 claims description 6
- OYEHPCDNVJXUIW-YPZZEJLDSA-N plutonium-242 Chemical compound [242Pu] OYEHPCDNVJXUIW-YPZZEJLDSA-N 0.000 claims description 5
- 239000000203 mixture Substances 0.000 claims description 3
- 230000000694 effects Effects 0.000 abstract description 3
- 239000000243 solution Substances 0.000 description 80
- 239000000463 material Substances 0.000 description 4
- 238000012958 reprocessing Methods 0.000 description 4
- 238000004088 simulation Methods 0.000 description 4
- 238000010586 diagram Methods 0.000 description 3
- 238000012986 modification Methods 0.000 description 3
- 230000004048 modification Effects 0.000 description 3
- -1 plutonium metal Chemical class 0.000 description 3
- 239000004698 Polyethylene Substances 0.000 description 2
- 229910052770 Uranium Inorganic materials 0.000 description 2
- 238000004422 calculation algorithm Methods 0.000 description 2
- 229920000573 polyethylene Polymers 0.000 description 2
- 238000005070 sampling Methods 0.000 description 2
- 239000002915 spent fuel radioactive waste Substances 0.000 description 2
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 0.000 description 2
- 238000004458 analytical method Methods 0.000 description 1
- 238000003556 assay Methods 0.000 description 1
- SHZGCJCMOBCMKK-KGJVWPDLSA-N beta-L-fucose Chemical compound C[C@@H]1O[C@H](O)[C@@H](O)[C@H](O)[C@@H]1O SHZGCJCMOBCMKK-KGJVWPDLSA-N 0.000 description 1
- 230000008033 biological extinction Effects 0.000 description 1
- 229910052793 cadmium Inorganic materials 0.000 description 1
- BDOSMKKIYDKNTQ-UHFFFAOYSA-N cadmium atom Chemical compound [Cd] BDOSMKKIYDKNTQ-UHFFFAOYSA-N 0.000 description 1
- 238000005516 engineering process Methods 0.000 description 1
- 230000001747 exhibiting effect Effects 0.000 description 1
- 238000011835 investigation Methods 0.000 description 1
- 238000000691 measurement method Methods 0.000 description 1
- 229910052751 metal Inorganic materials 0.000 description 1
- 239000002184 metal Substances 0.000 description 1
- 239000011259 mixed solution Substances 0.000 description 1
- ZQPKENGPMDNVKK-UHFFFAOYSA-N nitric acid;plutonium Chemical compound [Pu].O[N+]([O-])=O ZQPKENGPMDNVKK-UHFFFAOYSA-N 0.000 description 1
- 239000011824 nuclear material Substances 0.000 description 1
- 238000004886 process control Methods 0.000 description 1
- 238000011160 research Methods 0.000 description 1
- 229910001220 stainless steel Inorganic materials 0.000 description 1
- 239000010935 stainless steel Substances 0.000 description 1
- 229910002007 uranyl nitrate Inorganic materials 0.000 description 1
- 238000012795 verification Methods 0.000 description 1
Classifications
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- G—PHYSICS
- G01—MEASURING; TESTING
- G01T—MEASUREMENT OF NUCLEAR OR X-RADIATION
- G01T3/00—Measuring neutron radiation
- G01T3/001—Spectrometry
-
- G—PHYSICS
- G01—MEASURING; TESTING
- G01T—MEASUREMENT OF NUCLEAR OR X-RADIATION
- G01T3/00—Measuring neutron radiation
-
- G—PHYSICS
- G01—MEASURING; TESTING
- G01T—MEASUREMENT OF NUCLEAR OR X-RADIATION
- G01T1/00—Measuring X-radiation, gamma radiation, corpuscular radiation, or cosmic radiation
- G01T1/16—Measuring radiation intensity
- G01T1/17—Circuit arrangements not adapted to a particular type of detector
- G01T1/172—Circuit arrangements not adapted to a particular type of detector with coincidence circuit arrangements
-
- G—PHYSICS
- G01—MEASURING; TESTING
- G01T—MEASUREMENT OF NUCLEAR OR X-RADIATION
- G01T1/00—Measuring X-radiation, gamma radiation, corpuscular radiation, or cosmic radiation
- G01T1/16—Measuring radiation intensity
- G01T1/17—Circuit arrangements not adapted to a particular type of detector
- G01T1/178—Circuit arrangements not adapted to a particular type of detector for measuring specific activity in the presence of other radioactive substances, e.g. natural, in the air or in liquids such as rain water
-
- G—PHYSICS
- G01—MEASURING; TESTING
- G01T—MEASUREMENT OF NUCLEAR OR X-RADIATION
- G01T3/00—Measuring neutron radiation
- G01T3/008—Measuring neutron radiation using an ionisation chamber filled with a gas, liquid or solid, e.g. frozen liquid, dielectric
-
- G—PHYSICS
- G01—MEASURING; TESTING
- G01V—GEOPHYSICS; GRAVITATIONAL MEASUREMENTS; DETECTING MASSES OR OBJECTS; TAGS
- G01V5/00—Prospecting or detecting by the use of ionising radiation, e.g. of natural or induced radioactivity
- G01V5/20—Detecting prohibited goods, e.g. weapons, explosives, hazardous substances, contraband or smuggled objects
- G01V5/281—Detecting prohibited goods, e.g. weapons, explosives, hazardous substances, contraband or smuggled objects detecting special nuclear material [SNM], e.g. Uranium-235, Uranium-233 or Plutonium-239
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- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
- Y02E30/30—Nuclear fission reactors
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- Life Sciences & Earth Sciences (AREA)
- General Physics & Mathematics (AREA)
- High Energy & Nuclear Physics (AREA)
- Spectroscopy & Molecular Physics (AREA)
- Health & Medical Sciences (AREA)
- Molecular Biology (AREA)
- General Life Sciences & Earth Sciences (AREA)
- Geophysics (AREA)
- Measurement Of Radiation (AREA)
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Abstract
An estimation method and a monitoring system for a plutonium concentration in a uranium plutonium solution system is based on neutron coincidence counting. Corrections are performed of the neutron multiplication absorption effect of the spectrum difference between source neutrons and spontaneous fission source neutrons, and of the detection efficiency difference between induced fission neutrons and spontaneous fission source neutrons on a point model equation set. The calculated results of neutron coincidence counting by an improved point model equation set can accurately reflect a measurement result of neutron coincidence counting of a neutron detection system outside the uranium plutonium solution system.
Description
ESTIMATION METHOD AND MONITORING SYSTEM FOR PLUTONIUM CONCENTRATION IN URANIUM PLUTONIUM SOLUTION SYSTEM BASED ON NEUTRON COINCIDENCE COUNTING
Cross-Reference to Related Applications
The present application claims priority of invention patent application No 2020104115993 filed with China National Intellectual Property Administration on May 15, 2020, an entirety of which is incorporated herein by reference
Technical Field
The present invention belongs to a technology for monitoring and analyzing contents of fissile materials outside a reactor, is applicable to a field of estimating and monitoring a plutonium concentration in a uranium plutonium solution system, and particularly relates to an estimation method and a monitoring system for the plutonium concentration in the uranium plutonium solution system based on neutron coincidence counting
Background Art
During neutron detection of a uranium plutonium solution system, a plutonium concentration in the uranium plutonium solution system can be calculated or reversely derived by virtue of a total neutron counting rate and a coincidence neutron counting rate measured by a neutron detector, which method is called a NonDestructive Assay (NDA) method and does not need additional neutron sources.
A "point model" equation set derived by Boehnel (refer to Research on Application of Neutron Coincidence Counting in Attribute Determination of Plutonium. Shi Xueming, Liu Chengan, Nuclear Physics Review, 2004. 9) has been used as a main algorithm in plutonium concentration estimation: S = evsngs27°Ta24oM(1 ± (1) D = £2 -Vg2f2 gsrinnorD m2 [I_ ± (1+ a)(M vsf(rvid2 (2) Vsf2,Vieir -1) wherein S represents the total neutron counting rate, D represents the coincidence neutron counting rate, c represents a detection efficiency, gs273 represents a spontaneous fission rate of 240 n1240 represents a mass of 240Pu -e ffe ct ive, Ni represents a neutron leakage multiplication factor, i e, a ratio of the number of leakage neutrons eventually leaked out of the uranium plutonium solution system to the number of source neutrons produced in the uranium plutonium solution system after multiplication and absorption of spontaneous fission source neutrons and (a, n) source neutrons in the uranium plutonium solution system, and a represents a ratio of the number of (a, n) source neutrons to the number of spontaneous fission source neutrons. vsfd and 1/5/2 represent first and second moments for multiplicity distribution of spontaneous fission source neutrons of 240Pu yin and 1/id2 represent first and second moments for multiplicity distribution of induced fission neutrons fd, is a gate utilization factor.
A time coincidence measurement method is used in neutron coincidence counting, and measured objects are coincidence events, i.e., two or more events that occur simultaneously or that occur at moments associated with each other.
For a plutonium sample, multiplicity means that a plurality of indistinguishable neutrons are almost simultaneously released during spontaneous fission of plutonium, the number of neutrons released each time is random but follows a statistical law, and a distribution of the number of neutrons released each time is called as neutron multiplicity distribution The "point model" equation set is used to calculate neutron multiplicity based on the following assumptions: (1) all induced fission neutrons are released almost simultaneously with spontaneous fission source neutrons and (a, n) source neutrons, and a length of a fission chain is not considered; (2) neutron detection efficiency and fission probability are volumetrically uniform within the sample; (3) spontaneous fission source neutrons and (a, n) source neutrons have the same energy spectrum, and therefore, detection efficiency, fission probability, and induced fission multiplication are all the same, (4) a probability of neutron non-fissile capture is negligible; (5) neutron multiplicity and neutron energy are uncorrelated; and (6) time of extinction of neutrons in the sample/detector is not considered. These assumptions are basically tenable for a plutonium metal or plutonium oxide sample of a small volume, and the total neutron counting rate S and the coincidence neutron counting rate D calculated by the "point model" equation set is well coincident with calculated results through a three-dimensional Monte Carlo program simulation.
However, these assumptions are not always suitable for the uranium plutonium solution system. A lot of neutron moderator materials and neutron absorption materials are provided in the uranium plutonium solution system, and thus, the probability of neutron non-fissile capture can not be neglected. There is a large difference in energy spectra between spontaneous fission source neutrons and (a, n) source neutrons (as shown in FIG 1), and thus a difference also exists in the induced fission multiplication.
In addition, for the uranium plutonium solution system of a large volume, since induced fission occurs in a volumetrically nonuniform manner, the detection efficiencies for spontaneous fission source neutrons and induced fission neutrons differ.
In a word, the uranium plutonium solution system greatly differs from basic assumptions of the "point model" equation set; and if the equation set is directly adopted to estimate the plutonium concentration in the uranium plutonium solution system, a large deviation will be produced in the calculated results.
From an investigation of domestic and foreign applications of a neutron coincidence counting method in a plutonium mass monitoring field, no researcher has given an estimation method for a plutonium mass/plutonium concentration of a uranium plutonium solution system based on Boehnel's "point model" equation set From the viewpoint of engineering applications, due to a complex process flow, strong radioactivity and high critical safety requirements of a spent fuel reprocessing plant, timely and effective process control and safety monitoring are necessary during operation of the reprocessing plant. It is necessary to monitor the plutonium mass/plutonium concentration in the uranium plutonium solution system at key locations in the reprocessing plant.
Summary
An object of the present invention is to provide an estimation method and a monitoring system for a plutonium concentration in a uranium plutonium solution system based on neutron coincidence counting, which can meet requirements in key process flows of a spent fuel reprocessing plant for online monitoring of the plutonium concentration in the uranium plutonium solution system.
The present invention provides an estimation method for a plutonium concentration in a uranium plutonium solution system based on neutron coincidence counting. The estimation method includes: establishing a three-dimensional calculation model for the uranium plutonium solution system to calculate characteristic parameters when only (a, n) source neutrons are considered and when only spontaneous fission source neutrons are considered; calculating a detection efficiency Ei of a neutron detection system outside the uranium plutonium solution system for leakage neutrons when only spontaneous fission source neutrons are considered and a gate utilization factor f"), of the neutron detection system when only spontaneous fission source neutrons are considered, and concurrently calculating an energy spectrum of leakage neutrons when only spontaneous fission source neutrons are considered as a reference spectrum for calibrating the detection efficiency of the neutron detection system; determining correction factors, including a spectrum difference correction factor p for (a, n) source neutrons and spontaneous fission source neutrons and a detection efficiency difference correction factor t for induced fission neutrons and spontaneous fission source neutrons, wherein the spectrum difference correction factor p is a ratio of a detection efficiency E2 of the neutron detection system for leakage neutrons when only (a, n) source neutrons are considered to the detection efficiency Ei of the neutron detection system for leakage neutrons when only spontaneous fission source neutrons are considered, i.e., p = e2/e1, and the detection efficiency difference correction factor t is a ratio of a detection efficiency Eid of the neutron detection system for induced fission neutrons leaked out of the uranium plutonium solution system to the detection efficiency c1 of the neutron detection system for leakage neutrons when only spontaneous fission source neutrons are considered, i.e., t = sid/ei; substituting the calculated characteristic parameters, detection efficiency el, gate utilization factor fp and the determined correction factors p and t into an improved point model equation set to correct a measurement result of neutron coincidence counting of the neutron detection system; and obtaining a mass M240 of 240Pu-effective, by iteration, and estimating the plutonium concentration in the uranium plutonium solution system in combination with a volume of the uranium plutonium solution system.
In the estimation method, the improved point model equation set is: S = clysf1Ssf '1'240(41 aP M2) (3) "240"," , 2 f An 2 sf '1'240'1 JD'"1 - 2 ( Me2 -1 11 t2 D = Vsf2 V41171(12. vsfiavid2't 2 -1[1 (4) in which: S represents a total neutron counting rate; D represents a coincidence neutron counting rate; gsf represents a spontaneous fission rate of plutonium-240; ad represents the detection efficiency of the neutron detection system for leakage neutrons when only spontaneous fission source neutrons are considered; M1 represents a leakage multiplication coefficient when only spontaneous fission source neutrons are considered; M1 represents a net multiplication coefficient when only spontaneous fission source neutrons are considered; M2 represents a leakage multiplication coefficient when only (a, n) source neutrons are considered; M12 represents a net multiplication coefficient when only (a, n) source neutrons are considered; a represents a ratio of the number of (a, n) source neutrons to the number of spontaneous fission source neutrons; p represents the spectrum difference correction factor for (a, n) source neutrons and spontaneous fission source neutrons, t represents the detection efficiency difference correction factor for induced fission neutrons and spontaneous fission source neutrons; vsf, and 1sf2 represent first and second moments for multiplicity distribution of spontaneous fission neutrons of plutonium-240; vtdd and vid2 represent first and second moments for multiplicity distribution of induced fission neutrons caused by spontaneous fission source neutrons in the uranium plutonium solution system; Vidl' and vid2' represent first and second moments for multiplicity distribution of induced fission neutrons caused by (a, n) source neutrons in the uranium plutonium solution system; fp is the gate utilization factor of the neutron detection system, representing a ratio of coincidence neutron counting rates in gate durations tg and co, respectively; and m240 represents the mass of 240Pu-effective In the estimation method, by virtue of the calculated M240 as well as a plutonium isotope composition ratio and the volume of the uranium plutonium solution system, masses and concentrations of plutonium isotopes plutonium-238, plutonium-240 and plutonium-242 are further obtained by the equation (5): M240 = 2.52 X 238PU 240PU + 1.68 X 242PU (5) wherein 238pu, 240pu an,a 242 --Pu are masses of plutonium-238, plutonium-240 and plutonium-242, respectively.
In the estimation method, the leakage multiplication coefficient M1 when only spontaneous fission source neutrons are considered or the leakage multiplication coefficient M2 when only (a, n) source neutrons are considered is defined as a ratio of the number of leakage neutrons to the number of source neutrons after multiplication and absorption of source neutrons, which is calculated by: 1-Pid1 Or 2 -13C1 (Jr 2 (6) Al i or 2 n --rid, Or 21/id' 1 Or 2 The net multiplication coefficient Nil when only spontaneous fission source neutrons are considered or the net multiplication coefficient.02 when only (a, n) source neutrons are considered is defined as a ratio of the number of produced neutrons to the number of source neutrons after multiplication and absorption of source neutrons, which is calculated by: 1-1:4(11 Or 2 M or 2 = (7) -n-rid' Or 2ViC111 Or 2 wherein pi,' is a probability of causing induced fission by one source neutron, pc is a probability of capture (excluding fission capture), and \rich_ is the first moment for multiplicity distribution of induced fission neutrons, all of which can be calculated through a three-dimensional Monte Carlo program. The subscript 1 or 2 for each parameter indicates the case where only spontaneous fission source neutrons are considered or only (a, n) source neutrons are considered.
In the estimation method, the first and second moments Vidi and Vid2 for multiplicity distribution of induced fission neutrons caused by spontaneous fission source neutrons in the uranium plutonium solution system, the first and second moments Vidi' and id2' for multiplicity distribution of induced fission neutrons caused by (a, n) source neutrons in the uranium plutonium solution system, the detection efficiency E, of the neutron detection system for leakage neutrons when only spontaneous fission source neutrons are considered, the gate utilization factor fip, the spectrum difference correction factor p for (a, n) source neutrons and spontaneous fission source neutrons, and the detection efficiency difference correction factor t for induced fission neutrons and spontaneous fission source neutrons are calculated under a typical concentration of a uranium plutonium solution with a typical isotope mass ratio In the estimation method, the detection efficiency E, is estimated by a neutron reactivity corrected with the leakage multiplication coefficient M1, and is calculated by: Li = mi wherein FM, represents an average reactivity within a volume V of the neutron detection system when only spontaneous fission source neutrons are considered.
The gate utilization factor ID is calculated by: ID = Dg/D. (9) wherein Dg represents a coincidence neutron counting rate in a gate duration tg, and Don represents a coincidence neutron counting rate in a gate duration Go.
In the estimation method, El, E2 and 11d are calculated by: Fmixv Fm2xv Fmidxv -; and Eid = M2 Mid wherein FM, represents an average reactivity within a volume V of the neutron detection system when only spontaneous fission source neutrons are considered; FM2 represents an average reactivity within the volume V of the neutron detection system when only (a, n) source neutrons are considered; FR,' is an average reactivity of induced fission neutrons within the volume V of the neutron detection system, and Mid is a leakage multiplication coefficient of 88/11 xV (8) induced fission neutrons.
The present invention further provides a monitoring system for a plutonium concentration in a uranium plutonium solution system based on neutron coincidence counting, wherein the monitoring system includes a data analysis workstation and a plurality of neutron monitoring channels, wherein the data analysis workstation includes a data collection module, a data analysis and result output module and an alarm module, and wherein the data analysis and result output module includes a data analysis processing software which analyzes and processes measurement data of neutron coincidence counting using the estimation method as described above.
In the monitoring system, each of the neutron monitoring channels includes a neutron detection assembly, a neutron coincidence circuit and an in-situ display device, wherein original pulse signals detected by the neutron detection assembly are processed in real time by the neutron coincidence circuit to form the measurement data of neutron coincidence counting, the measurement data is displayed in real time on the in-situ display device and transmitted to the data analysis workstation, collected and processed by the data collection module, and then calculated and analyzed by the data analysis processing software, and wherein the neutron detection assembly includes a neutron detector as well as a moderator and a shield adapted for the neutron detector.
In the monitoring system, the data collection module includes a plurality of data collection channels.
In the monitoring system, the alarm module sends out an alarm signal when the data analysis processing software determines that the plutonium concentration in the uranium plutonium solution system exceeds a preset threshold, The present invention has the following advantageous effects: The estimation method and the monitoring system for the plutonium concentration in the uranium plutonium solution system based on neutron coincidence counting of the present invention perform corrections of the spectrum difference between (a, n) source neutrons and spontaneous fission source neutrons, and of the detection efficiency difference between induced fission neutrons and spontaneous fission source neutrons, on the point model equation set, so that the calculated results of neutron coincidence counting by the improved point model equation set can accurately reflect the measurement result of neutron coincidence counting of the neutron detection system outside the uranium plutonium solution system, thereby accurately estimating the plutonium concentration in the uranium plutonium solution system. This estimation method realizes nondestructive monitoring and analysis of the plutonium concentration in the uranium plutonium solution system, and is an advanced estimation method for the plutonium concentration in the uranium plutonium solution system with engineering feasibility.
The present invention effectively solves a problem that an original point model equation set is not suitable for the uranium plutonium solution system. In view of differences between the basic assumptions of the point model equation set and the uranium plutonium solution system, a net multiplication coefficient is introduced to correct neutron multiplication and absorption effect, a spectrum difference correction factor for (a, n) source neutrons and spontaneous fission source neutrons is introduced to correct the spectrum difference between (a, n) source neutrons and spontaneous fission source neutrons, and a detection efficiency difference correction factor for induced fission neutrons and spontaneous fission source neutrons is introduced to correct the detection efficiency difference between induced fission neutrons and spontaneous fission source neutrons, so that a predicted result of the coincidence neutron counting rate obtained by the improved point model equation set is substantially consistent with the calculated results through the three-dimensional Monte Carlo program simulation. The method is applicable to estimation of the plutonium mass/plutonium concentration in the uranium plutonium solution system, and thus solves a key problem of neutron coincidence counting in estimation of the plutonium concentration in the uranium plutonium solution system
Brief Description of the Drawings
FIG. 1 is a comparison diagram of energy spectra of spontaneous fission source neutrons and (ct, n) source neutrons generated by a typical uranium plutonium solution system per se; FIG. 2 is a layout diagram of an embodiment of an estimation method for a plutonium concentration in a uranium plutonium solution system based on neutron coincidence counting; FIG. 3 is a flowchart of an embodiment of the estimation method for the plutonium concentration in the uranium plutonium solution system based on neutron coincidence counting; and FIG. 4 is a structural block diagram of an embodiment of a monitoring system for a plutonium concentration in a uranium plutonium solution system based on neutron coincidence counting.
Detailed Description of the Embodiments
Implementations of the present invention will be described in detail below with reference to accompanying drawings and embodiments.
Here, a solution tank 1 (as a solution device) containing a uranium plutonium solution is taken as an example to illustrate how the present invention is applied to estimate and monitor a plutonium concentration in a uranium plutonium solution system. The uranium plutonium solution system includes the solution tank 1 and the uranium plutonium solution contained in the solution tank 1. As shown in FIG. 2, the solution tank 1 is 10cm high, has an inner diameter of 12cm, and acts as a stainless steel container with a wall thickness of 3mm. The solution tank 1 is filled with a mixed solution of uranyl nitrate and plutonium nitrate (i.e., the uranium plutonium solution). Twenty four BF3 detectors 311 each coated with polyethylene 312 (as a moderator) are arranged around the solution tank 1, and a cadmium (Cd) shield layer 313 is covered outside the polyethylene 312. Typically, a neutron detection system is provided outside the uranium plutonium solution system. The neutron detection system is an integrated system of hardware and software structures which monitor the plutonium concentration in the uranium plutonium solution system. Here, the volume or capacity of the solution tank I and the number of detectors 311 are merely exemplary, and do not constitute any limitation of the present application.
In addition, the arrangement of the detectors 311 is not limited to that shown in FIG. 2.
As shown in FIG. 3, an estimation method for a plutonium concentration in a uranium plutonium solution system based on neutron coincidence counting according to the present invention includes: establishing a three-dimensional calculation model for the uranium plutonium solution system to calculate characteristic parameters when only (a, n) source neutrons are considered and when only spontaneous fission source neutrons are considered (Step 100); calculating a detection efficiency El of a neutron detection system outside the uranium plutonium solution system for leakage neutrons when only spontaneous fission source neutrons are considered and a gate utilization factor f of the neutron detection system when only spontaneous fission source neutrons are considered, and concurrently calculating an energy spectrum of leakage neutrons when only spontaneous fission source neutrons are considered as a reference spectrum for calibrating the detection efficiency of the neutron detection system (Step 200); determining correction factors, including a spectrum difference correction factor p for (a, n) source neutrons and spontaneous fission source neutrons and a detection efficiency difference correction factor t for induced fission neutrons and spontaneous fission source neutrons, wherein the spectrum difference correction factor p is a ratio of a detection efficiency E2 of the neutron detection system for leakage neutrons when only (a, n) source neutrons are considered to the detection efficiency El of the neutron detection system for leakage neutrons when only spontaneous fission source neutrons are considered, i.e., p = E2/E1, and the detection efficiency difference correction factor t is a ratio of a detection efficiency cid of the neutron detection system for induced fission neutrons leaked out of the uranium plutonium solution system to the detection efficiency El of the neutron detection system for leakage neutrons when only spontaneous fission source neutrons are considered, i.e., t = Eid/Ei (Step 300); substituting the calculated characteristic parameters, detection efficiency El, gate utilization factor ID and the determined correction factors p and t into an improved point model equation set to correct a measurement result of neutron coincidence counting of the neutron detection system (Step 400); and obtaining a mass ni240 of 240Pu-effective, by iteration, and estimating the plutonium concentration in the uranium plutonium solution system in combination with a volume of the uranium plutonium solution system (Step 500).
The improved point model equation set is: S = E1vsf19s2/1° Inzao(Mi + aPM2) (3) s f 1"240'12./Di'12 2 (M1 -1 M2 -1"] V s f2 ± V s f 1V id2t- 2 1) ± V s f laV ic12' ( 1)1 V idl V idl (4) in which: S represents a total neutron counting rate; D represents a coincidence neutron counting rate; gsf represents a spontaneous fission rate of plutonium-240; Ei represents the detection efficiency of the neutron detection system for leakage neutrons when only spontaneous fission source neutrons are considered; M1 represents a leakage multiplication coefficient when only spontaneous fission source neutrons are considered; /14*1 represents a net multiplication coefficient when only spontaneous fission source neutrons are considered; M2 represents a leakage multiplication coefficient when only (a, n) source neutrons are considered; NI2 represents a net multiplication coefficient when only (a, n) source neutrons are considered; a represents a ratio of the number of (a, n) source neutrons to the number of spontaneous fission source neutrons, p represents the spectrum difference correction factor for a, n source neutrons and spontaneous fission source neutrons; t represents the detection efficiency difference correction factor for induced fission neutrons and spontaneous fission source neutrons, Vsfl and vsf2 represent first and second moments for multiplicity distribution of spontaneous fission source neutrons of plutonium-240; vidi and vid2 represent first and second moments for multiplicity distribution of induced fission neutrons caused by spontaneous fission source neutrons in the uranium plutonium solution system; yid(' and vica' represent first and second moments for multiplicity distribution of induced fission neutrons caused by (a, n) source neutrons in the uranium plutonium solution system; fp is the gate utilization factor of the neutron detection system, representing a ratio of coincidence neutron counting rates in gate durations tg and cc, respectively; and 7n240 represents the mass of 240Pu-effective.
With reference to nuclear material manuals, the spontaneous fission rate 24f0 of 4 2 0 pu, gs and the first moment vsfi and the second moment vsf2 for multiplicity distribution of spontaneous fission source neutrons are fixed values 475.276 fission/(ps), 2.156 and 3.825, respectively.
Firstly, in Step 100, a three-dimensional calculation model is established for the uranium plutonium solution system using a three-dimensional Monte Carlo program to calculate characteristic parameters when only (a, n) source neutrons are considered and when only spontaneous fission source neutrons are considered. Here, the required characteristic parameters are calculated by taking one of a uranium concentration, a plutonium concentration, a uranium isotope mass ratio and a plutonium isotope mass ratio as an example, wherein the leakage multiplication coefficient NI is a ratio of the number of leakage neutrons to the number of source neutrons after multiplication and absorption of source neutrons, the net multiplication coefficient NE is a ratio of the number of produced neutrons to the number of source neutrons after multiplication and absorption of source neutrons, and these parameters can be calculated by the equations (6) and (7).
11/1 _ Or 2 1 or 2 ---vial or 2victr1 or 2 1-Pidl or 2 M 1 or 2 = (7) 1-Pid1 Or 2Vidll 0r2 phi, pc and yidi are respectively a probability of causing induced fission by one source neutron, a probability of capture (excluding fission capture), and the first moment for multiplicity distribution of induced fission neutrons, all of which can be calculated by the three-dimensional Monte Carlo program Mei, M2 and 02 can be calculated by the equations (6) and (7), and can also be obtained by calculation based on Monte Carlo program modeling.
The calculated results of these characteristic parameters are shown in Table I. TABLE 1 Calculated results of characteristic parameters Only spontaneous fission source Only (a, n) source neutrons are neutrons are considered considered Vial 2.8878 V ' 2.8951 idl Vid2 6.8804 V id2' 6.9214 Pid 0.0586 Pict' 0.0380 Pc 0.0878 Pc 0.0604 M1 1.0275 M2 L0130 NI 1.1331 142 L0809 Next, in Step 200, the three-dimensional Monte Carlo program is used to calculate the detection efficiency c of the neutron detection system for leakage neutrons when only spontaneous fission source neutrons are considered and the gate utilization factor fp of the neutron detection system when only spontaneous fission source neutrons are considered, and an energy spectrum of leakage neutrons when only spontaneous fission source neutrons are considered is concurrently calculated as a reference spectrum for calibrating the detection efficiency of the neutron detection system. The detection efficiency El is estimated by a neutron reactivity (e.g., a (n, a) reactivity for a BF3 detector) corrected with the leakage multiplication coefficient M1, and is calculated by the equation (8).
Et = FM, x V (8) The gate utilization factor fp is defined as a ratio of coincidence neutron counting rates in gate durations tg and GO, respectively, and is calculated by the equation (9).
= Dg/Do, (9) tg represents a time window in which the coincidence neutron counting is considered effective, which time window should neither be too short, or the detection efficiency will be reduced; nor too long, or occasional coincident neutron counting can not be distinguished effectively. Therefore, tg is a length of IS time determined according to actual circumstances of a specific application situation. An average reactivity within a volume V of the neutron detection system and the coincidence neutron counting rates for gate durations Eg and 00 can all be calculated by the three-dimensional Monte Carlo program. It is to be noted that all of the detection efficiency, the gate utilization factor and calibration requirements of the neutron detection system calculated here are directed to the case where only spontaneous fission source neutrons are considered. The calibration requirements of the neutron detection system can be met by calculating the energy spectrum of leakage neutrons leaked from the solution device (the uranium plutonium solution system) as a reference spectrum for calibrating the detection efficiency of the neutron detection system. For the above embodiment, E1 is calculated as: c = 1.59966%/1.02746 = 1.55691%; when cq = 64as, fp is calculated as: fp = 8.7373 x 10'5/2.9236x 10-4 = 0.2889.
In addition, in Step 300, it is also required to determine the spectrum difference correction factor p for (a, n) source neutrons and spontaneous fission source neutrons and the detection efficiency difference correction factor t for induced fission neutrons and spontaneous fission source neutrons.
First, El, E2 and Eid are calculated by: Emuixv Em2xv Emidxv 81= -2 Mid wherein FM, represents an average reactivity within the volume V of the neutron detection system when only spontaneous fission source neutrons are considered, FM2 represents an average reactivity within the volume V of the neutron detection system when only (a, n) source neutrons are considered; FMid is an average reactivity of induced fission neutrons within the volume V of the neutron detection system; and Mid is a leakage multiplication coefficient of induced fission neutrons.
Then, the spectrum difference correction factor p for (a, n) source neutrons and spontaneous fission source neutrons is calculated by the equation (10).
P = E2/E1 (10) Meanwhile, the detection efficiency difference correction factor t for induced fission neutrons and spontaneous fission source neutrons is calculated by the equation (11).
t = Eid/Ei (11) The detection efficiency El of the neutron detection system for leakage neutrons when only spontaneous fission source neutrons are considered and the detection efficiency E2 of the neutron detection system for leakage neutrons when only (a, n) source neutrons arc considered can be calculated by setting different neutron source items and referring to the equation (8). The detection efficiency Eid of the neutron detection system for induced fission neutrons leaked out of the uranium plutonium solution system is obtained by calculating a difference between results of turning on and off a switch for producing induced fission neutrons According to the above method, the following results are obtained: p =E2/ci, 1.54467%/1.55691% = 0.9921, and t = eid/ci = 1.61436%/1.55691% = 1.0369.
Next, in Step 400, the measurement result of neutron coincidence counting of the neutron detection system is corrected with the calculated characteristic parameters, detection efficiency c1, gate utilization factor to and the determined correction factors (especially the spectrum difference correction factor p for (a, n) source neutrons and spontaneous fission source neutrons and the detection efficiency difference correction factor t for induced fission neutrons and spontaneous fission source neutrons). Due to a lack of actual measurements of S and D, predicted results of the three-dimensional Monte Carlo program are used as counting results of the neutron detection system, and when tp = 64ps, S = 1500.62/s, D = 2.82/s. By substituting the calculated results of the above parameters such as the characteristic parameters, the 240 detection efficiency, the correction factors (v511, vs122, gsi and the calculated parameters al, fp, p, t,72 -idl, V1d2, 1)&1', V(d2') equation set including the equations (3) and (4), it is obtained by calculation that n1240 21.437g (refer to the equation (12)), which deviates from the mass 21.441g of 24"Pu-effecti ve adopted by the actual simulation of the three-dimensional Monte Carlo program by only 0.02%, exhibiting a high estimation accuracy.
7112.40 = 2[D Ei/Dmi2via2't2s A1.2-1)1 [vi 2pm2 -11" 11191d2 4-7 (M1 -fl /M2-1 t2 A41 ( (41 -1) p11421 9s210 Ei2 fp Ail 2 VS Vs fit.? id2 vidi-1 (12) Finally, in Step 500, the mass 171240 of 2411Pu-effective is obtained by iteration using the equations (3) and (4), and the plutonium concentration in the uranium plutonium solution system is estimated in combination with the volume of the solution device (i.e the volume of the uranium plutonium solution system).
into the improved point model In contrast, if the plutonium concentration is estimated by the original point model equation set including the equations (1) and (2), the characteristic parameters when spontaneous fission source neutrons and (a, n) source neutrons are simultaneously considered are: vidi = 2.8934, Vid2 = 6.9120, M = 1.0163, e = 1.5475, fp = 0.3004, and it is obtained that m240=38.673g (refer to the equation (13)), which deviates from the mass 21.441g of 240Pu-effective adopted by the actual simulation of the three-dimensional Monte Carlo program by up to 80%, indicating a great deviation in the estimated result.
M240 - 1)-Efnmsvi12(A1-1) (13) idi -1)E2 V sf 29.52rf D 11112 ID Further,by virtue of the calculated M240, as well as the plutonium isotope composition ratio and the volume of the uranium plutonium solution system that are known, masses and concentrations of plutonium isotopes plutonium-238, plutonium-240 and plutonium-242 can be further obtained using the equation (5).
m240 -2.52 X 238PU + 1.68 X 242PU (5) It is to be noted that when the plutonium concentration in the uranium plutonium solution system is estimated, the mass M240 of 240Pu-effective and the plutonium concentration are calculated using the equations (3) and (4), and at this time, a value of a is determined according to measured values of S and D; and when accuracy of the estimated results is evaluated, it can be done by directly substituting the estimated results into the equation (12) for verification, and at this time, there is no need to determine the value of a.
It is to be further noted that in the three-dimensional modeling and numerical simulation of the uranium plutonium solution system, the Monte Carlo program (a statistical simulation method based on random sampling) is taken as an example for description. However, implementing of technical solutions of the present invention does not depend on the selected numerical calculation program or algorithm. For example, the estimation method of the present invention can also be performed by a Las Vegas program which is also based on random sampling.
The present invention further provides a monitoring system 10 for a plutonium concentration in a uranium plutonium solution system based on neutron coincidence counting, which has a logic structure as shown in FIG. 4. The monitoring system 10 shown in FIG. 4 includes a data analysis workstation 20 and a plurality of neutron monitoring channels 30. The data analysis workstation 20 includes a data collection module 21, a data analysis and result output module 22 and an alarm module 23. The data analysis and result output module 22 includes a data analysis processing software which analyzes and processes measurement data of neutron coincidence counting using the estimation method as described above. Each of the neutron monitoring channels includes a neutron detection assembly 31, a neutron coincidence circuit 32 and an in-situ display device 33, as well as related mounting brackets, pipes, cables, shielding materials and the like. Original pulse signals detected by the neutron detection assembly 31 are processed in real time by the neutron coincidence circuit 32 to form the measurement data of neutron coincidence counting, and the measurement data is displayed in real time on the in-situ display device 33 and transmitted to the data analysis workstation 20, collected and processed by the data collection module 21, and then calculated and analyzed by the data analysis processing software. The neutron detection assembly 31 includes a neutron detector 311 as well as a moderator 312 and a shield 313 adapted for the neutron detector 311, and the specific structure is shown in FIG. 2. The data collection module 21 includes a plurality of data collection channels that can collect data for different neutron monitoring channels respectively. The alarm module 23 sends out an alarm signal when the data analysis processing software determines that the plutonium concentration in the uranium plutonium solution system exceeds a preset threshold.
Apparently, those skilled in the art can make various changes and modifications to the present invention without departing from the spirit and scope of the present invention. Thus, such changes and modifications to the present invention fall within the scopes of the claims of the present invention and equivalents thereof and the present invention is intended to include such changes and modifications as well
Claims (1)
- What is claimed is 1. An estimation method for a plutonium concentration in a uranium plutonium solution system based on neutron coincidence counting, the estimation method comprising: establishing a three-dimensional calculation model for the uranium plutonium solution system to calculate characteristic parameters when only (a, n) source neutrons are considered and when only spontaneous fission source neutrons are considered; calculating a detection efficiency El of a neutron detection system outside the uranium plutonium solution system for leakage neutrons when only spontaneous fission source neutrons are considered and a gate utilization factor fp of the neutron detection system when only spontaneous fission source neutrons are considered, and concurrently calculating an energy spectrum of leakage neutrons when only spontaneous fission source neutrons are considered as a reference spectrum for calibrating the detection efficiency of the neutron detection system; determining correction factors, including a spectrum difference correction factor p for (a, n) source neutrons and spontaneous fission source neutrons and a detection efficiency difference correction factor t for induced fission neutrons and spontaneous fission source neutrons, wherein the spectrum difference correction factor p is a ratio of a detection efficiency £2 of the neutron detection system for leakage neutrons when only (a, n) source neutrons are considered to the detection efficiency £, of the neutron detection system for leakage neutrons when only spontaneous fission source neutrons are considered, i e p = £2/a1, and the detection efficiency difference correction factor t is a ratio of a detection efficiency Eid of the neutron detection system for induced fission neutrons leaked out of the uranium plutonium solution system to the detection efficiency z of the neutron detection system for leakage neutrons when only spontaneous fission source neutrons are considered, i.e. t = Lid/El; substituting the calculated characteristic parameters, detection efficiency El, gate utilization factor fp and the determined correction factors p and t into an improved point model equation set to correct a measurement result of neutron coincidence counting of the neutron detection system; and obtaining a mass m240 of 240Pu-effective, by iteration, and estimating the plutonium concentration in the uranium plutonium solution system in combination with a volume of the uranium plutonium solution system 2. The estimation method according to claim 1, wherein the improved point model equation set is: n240 S = Eivsfissf m243(Mi+ riPM2) and efl°M240E12fDM12 Vsf a vid2 t2 MI2,1131 D - vsf2 Vsf Vic/2 t2 M 1 2 V idl in which: S represents a total neutron counting rate; D represents a coincidence neutron counting rate; gsf represents a spontaneous fission rate of plutonium-240; ci represents the detection efficiency of the neutron detection system for leakage neutrons when only spontaneous fission source neutrons are considered; M, represents a leakage multiplication coefficient when only spontaneous fission source neutrons are considered; MI, represents a net multiplication coefficient when only spontaneous fission source neutrons are considered; M2 represents a leakage multiplication coefficient when only (a, n) source neutrons are considered; M12 represents a net multiplication coefficient when only (a, n) source neutrons are considered; a represents a ratio of the number of (a, n) source neutrons to the number of spontaneous fission source neutrons; p represents the spectrum difference correction factor for a, n source neutrons and spontaneous fission source neutrons; t represents the detection efficiency difference correction factor for induced fission neutrons and spontaneous fission source neutrons; vsfi and vsf2 represent first and second moments for multiplicity distribution of spontaneous fission source neutrons of plutonium-240; vidi and v1d2 represent first and second moments for multiplicity distribution of induced fission neutrons caused by spontaneous fission source neutrons in the uranium plutonium solution system; yid,: and Vid2 represent first and second moments for multiplicity distribution of induced fission neutrons caused by (a, n) source neutrons in the uranium plutonium solution system; fp is the gate utilization factor of the neutron detection system, representing a ratio of coincidence neutron counting rates in gate durations tg and co, respectively; and M240 represents the mass of 240Pu-effective.3. The estimation method according to claim 2, wherein the leakage multiplication coefficient M, when only spontaneous fission source neutrons are considered or the leakage multiplication coefficient M2 when only (a, n) source neutrons are considered is defined as a ratio of the number of leakage neutrons to the number of source neutrons after multiplication and absorption of source neutrons, which is calculated by: M1 or 2 = tPidj or 2 -Prl or 2 OT 2Vidll Or 2 the net multiplication coefficient 01 when only spontaneous fission source neutrons are considered or the net multiplication coefficient M*2 when only (a, n) source neutrons are considered is defined as a ratio of the number of produced neutrons to the number of source neutrons after multiplication and absorption of source neutrons, which is calculated by: or 2 - 1-Pid1 Or 2 An' n' 1 wherein pi,' is a probability of causing induced fission by one source neutron, pc is a probability of capture, and yid, is the first moment for multiplicity distribution of induced fission neutrons 4. The estimation method according to claim 2, wherein the first and second moments Vidi and Vid2 for multiplicity distribution of induced fission neutrons caused by spontaneous fission source neutrons in the uranium plutonium solution system, the first and second moments id,' and vid2' for multiplicity distribution of induced fission neutrons caused by (cc, n) source neutrons in the uranium plutonium solution system, the detection efficiency E, of the neutron detection system for leakage neutrons when only spontaneous fission source neutrons are considered, the gate utilization factor fp, the spectrum difference correction factor p for (a, n) source neutrons and spontaneous fission source neutrons, and the detection efficiency difference correction factor t for induced fission neutrons and spontaneous fission source neutrons are calculated under a typical concentration of a uranium plutonium solution with a typical isotope mass ratio.5. The estimation method according to claim 4, wherein the detection efficiency c, is estimated by a neutron reactivity corrected with the leakage multiplication coefficient M1, and is calculated by: Fm,xv Cl = wherein FM, represents an average reactivity within a volume V of the neutron detection system when only spontaneous fission source neutrons are considered; the gate utilization factor ID is calculated by: 1-Pid1 or 2Vidl1 Or 2 wherein Dg represents a coincidence neutron counting rate in a gate duration tg, and Dee represents a coincidence neutron counting rate in a gate duration Go.6. The estimation method according to claim 4, wherein cid are calculated by: *FM2xV, and 11d= FMIdXV 1142 mtd ' El 12 and E2 = FM, x V. £1 =Mwherein FM, represents an average reactivity within a volume V of the neutron detection system when only spontaneous fission source neutrons are considered; FM2 represents an average reactivity within the volume V of the neutron detection system when only (cc, n) source neutrons are considered; FMi, is an average reactivity of induced fission neutrons within the volume V of the neutron detection system; and Mid is a leakage multiplication coefficient of induced fission neutrons.7. The estimation method according to claim 2, wherein by virtue of the calculated 7/2240 as well as a plutonium isotope composition ratio and the volume of the uranium plutonium solution system, masses and concentrations of plutonium isotopes plutonium-238, plutonium-240 and plutonium-242 are further obtained by: m2cco = 2.52 x 238Pu + 240Pu + 1.68 x 242PU wherein 238Pu, 24"Pu and 242Pu are masses of plutonium-238, plutonium-240 and plutonium-242, respectively.8. A monitoring system for a plutonium concentration in a uranium plutonium solution system based on neutron coincidence counting, wherein the monitoring system comprises a data analysis workstation and a plurality of neutron monitoring channels, wherein the data analysis workstation includes a data collection module, a data analysis and result output module and an alarm module, and wherein the data analysis and result output module includes a data analysis processing software which analyzes and processes measurement data of neutron coincidence counting using the estimation method according to any one of claims 1 to 7.9. The monitoring system according to claim 8, wherein each of the neutron monitoring channels includes a neutron detection assembly, a neutron coincidence circuit and an in-situ display device, wherein original pulse signals detected by the neutron detection assembly are processed in real time by the neutron coincidence circuit to form the measurement data of neutron coincidence counting, the measurement data is displayed in real time on the in-situ display device and transmitted to the data analysis workstation, collected and processed by the data collection module, and then calculated and analyzed by the data analysis processing software, and wherein the neutron detection assembly includes a neutron detector as well as a moderator and a shield adapted for the neutron detector.10. The monitoring system according to claim 9, wherein the data collection module includes a plurality of data collection channels 11. The monitoring system according to claim 9, wherein the alarm module sends out an alarm signal when the data analysis processing software determines that the plutonium concentration in the uranium plutonium solution system exceeds a preset threshold.
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