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Annual Report of China Institute of Atomic Energy 2005
Nuclear Safeguards Techniques
1
2005 China-U.S. Joint Integrated Nuclear Material Management Demonstration
LIU Da-ming，GAN Lin
The 2005 Joint China-U.S. Integrated Nuclear Material Management Technology Demonstration was
opened on October 24 at China Institute of Atomic Energy(CIAE) in Beijing，China. Mr. Sun Qin，
chairman of China Atomic Energy Authority (CAEA) and Ambassador Linton Brooks，administrator of
the Department of Energy’s(DOE) National Nuclear Security Administration(NNSA) presided over the
opening ceremony of this demonstration. Representatives from both sides of governmental agencies and
China National Nuclear Corporation(CNNC) toured the facility. Approximately 80 representatives from
various Chinese nuclear facilities and institutes participated in hands-on training，seminars and exercises
throughout the week.
The purpose of the demonstration is to promote the understanding and adoption of modern security
concepts and technologies at nuclear facilities by demonstrating physical protection，nuclear material
control and accounting，and international safeguards technologies which is as a first line of defense
against nuclear material theft，diversion and sabotage. The activities for the demonstration included joint
technical work on hardware and software，system design and installation，and topical workshops on
vulnerability assessment and nondestructive assay，exchange of technical personnel，training，and site
visits.
CAEA and NNSA are the government sponsors of the event，which is being conducted under the
auspices of the China-U.S. Peaceful Uses of Nuclear Technology Agreement(PUNT). Experts from the
CIAE，the Fourth Institute of Nuclear Engineering of CNNC，and four US DOE national laboratories (Los
Alamos，Lawrence，Livermore，Sandia and Oak Ridge National Laboratories) are involved to provide the
technical expertise to support the planning and implementation.
The follow-up cooperative activities related to nuclear material safeguards will be implemented
further. A near term project is a regional workshop on “Facility Systems of Nuclear Material Accounting
and Control at Facilities” in cooperated with the International Atomic Energy Agency(IAEA) will be held
in the May of 2006.
2 Improvements on Nuclear Power Plant Burnup Measurement System and
Burnup Measurements of Daya Bay Spent Fuel Assemblies
DONG Ming-li，LU Wen-guang，LIU Guo-liang1，LIU En-peng1，LU Xiao-qing1
（1 Lanzhou Nuclear Fuel Complex，Lanzhou 732850，China）
The fork detector spent fuel measurement system has been improved. The improved system still
includes the gross gamma，the high-resolution gamma spectrometry(HRGS) and the passive neutron
measurements.
FUNDAMENTAL AND APPLIED FUNDAMENTAL RESEARCH·Nuclear Safeguards Techniques
175
1) In order to get spent fuel homogeneous measurement，the geometry of the detector body has been
improved，so mean burnup has been measured，and the accuracy of measured burnup becomes better.
2) A higher accuracy collimator of the HRGS was used，hence the wider range of activity intensity
can be measured and the reliable data of the HRGS have been got.
3) The passive neutron energy spectrometry and the electric disturbance signal spectrometry have
been measured. The strip off the electric disturbance signal and background spectrometry method was
used，so that it can get the better results under the more intensity of electric disturbance signal.
4) According to the measured data of the HRGS and the passive neutron，respectively，the modular
calculations of
235
U initial enrichments，burnup，total plutonium and total fissile isotope content were
achieved by using nervous web. The calculation results demonstrate that the describing function
relationship between the measured parameter and initial enrichments，burnup，total plutonium and total
fissile isotope content by using the nervous web modular calculations is better than that mathematics
formula description. The modular calculations of
235U
initial enrichments are different from the abroad
active neutron measurement，especially. The modular calculations of
235U
initial enrichments not only
save on an expensive active neutron source，a heavy shield and a complicated operation system，but also
improve the accuracy.
Nineteen PWR assemblies of the Daya Bay nuclear power plant have been measured. The initial
enrichments of the assemblies measured are 1.80%，2.40% and 3.10%. The range of burnup average value
in the assemblies is 14 to 33 GW·d/t. The cooling time varies from 8 to 9 a.
The burnup profiles were obtained by measuring the gross gamma at each axial location of the spent
fuel assemblies. The burnup，total plutonium content and total fissile isotope content were obtained by
measuring the passive neutron counting. The errors of the burnup，total plutonium content and total fissile
isotope content are better than 1.6%，1.2% and 1.8%，respectively. The burnup and the cooling time were
obtained by measuring the HRGS. The errors of the burnup and he cooling time are better than 5.0%，
respectively. The 235U initial enrichment，burnup，total plutonium content and total fissile isotope content
were obtained by using nervous web modular calculations. The errors of the
235U
initial enrichments，
burnup，total plutonium content and total fissile isotope content are better than 1.0%，3.0%，2.0% and
2.7%，respectively.
The above errors of the measured 235U initial enrichments，burnup，total plutonium content and total
fissile isotope content are much same as，even better than errors measured at home and abroad.
3 Locating and Transferring of Uranium-Bearing Particles in Swipe Samples
by Fission Track and Micromanipulation
SHEN Yan，ZHAO Yong-gang，GUO Shi-lun
Fission track and micromanipulation technique is an effective method for locating and transferring of
the uranium-bearing particles in swipe samples. It is very important especially for sparse samples. So this
technique was studied and established in our laboratory. The method includes two steps as follows：
locating the particles containing uranium by fission track，and transferring the particles with the assistance
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Annual Report of China Institute of Atomic Energy 2005
of a optical microscope equipped with micromanipulator.
Firstly，the particles in swipe sample were recovered by a piece of polycarbonate membrane. The
membrane was coated with a piece of mica slice as track detector，and irradiated with thermal neutron，the
－
neutron fluence is about 1015-1016 cm 2. After irradiation，the fission track detector was etched with HF
acid at 25 ℃ for 15 min. Secondly，the uranium particles were located through the fission track and picked
up with a glass micro needle under microscope. The particles can be transferred either onto rhenium belt
for TIMS or onto graphite disk for SIMS determination.
4
Development of Integrated Total Neutron Measurement System
ZHU Li-qun，LIU Guo-rong，WANG Xiao-zhong，XU Xiao-ming，SHEN Ning，LU Xue-sheng，
BAI Lei，HUANG Ping，MENG Yan-tai，PAN Da-jin，JIA Xiang-jun
An integrated total neutron measurement system has been developed. The data acquisition software
system and miniaturized electronics components have taken place of the original neutron assay instrument.
All of electronics components are fixed in the head of the detector，which are made up of high voltage
converter，signal processor and data collection circuit (SCM CPU and its periphery circuit) etc. It can be
operated independently with the LED which can freely rotate about 180º or with a computer.
The system has a small light body，so it can be moved conveniently. It has a special measurement
direction. This system can be used in the nuclear inspection and verification，the facility storage
monitoring and also the nuclear waste searching.
The integrated total neutron measurement system has been tested for a long time，and the stability of
the performances is reliability.
5 Development of Vehicle Security Portal
for Special Nuclear Material and Other Radionuclides
ZHAO Rong-sheng，ZHANG Wen-liang，LI Xin-jun，FANG Xin，ZHANG Li-ping
After indoor debugging the vehicle portal special nuclear material monitor was mounted at the
outdoor scene in 2005. The tests of detection sensitivity of the monitor for high enrichment uranium(HEU)
and low-burnup plutonium with different type of vehicles have been performed. The tests of statistical
false alarm rate and stability of the monitor have been performed too.
The results of the tests are as follows：the detection sensitivity is < 192 g for HEU，and <10 g for
－
low-burnup plutonium；the statistical false alarm rate of the monitor is 7.68×10 6.
Above performances of the monitor are much better than the prearranged performances and the
performances of overseas congener.
The monitor has been operating continuously for ten months and it is proved it has reliable
performance. In general，a conclusion has been drawn that the monitor can be used for detection of special
nuclear material and radiation.
FUNDAMENTAL AND APPLIED FUNDAMENTAL RESEARCH·Nuclear Safeguards Techniques
6
177
Development of Active Well Neutron Coincidence Counting System
WANG Xiao-zhong，JIA Xiang-jun，MENG Yan-tai，LIU Da-ming，SUNG Shao-fei，
XU Xiao-ming，ZHU Li-qun，BAI Lei，HUANG Ping，GAN Lin
The active well neutron coincidence counter(AWCC) has been developed，which is one of the main
devices on non-destructive assay for the U-bearing nuclear material. The AWCC bases on the thermal
neutron induced fission of 235U，and the induced fission neutron source in the device is two Am-Li neutron
sources with low energy neutron emitting.
The shape of AWCC likes a column with a hole in the center. There are two pieces of plug in center
hole for 4π measuring condition. 42 3He counters were embedded in the polythene，which were divided
into 6 groups and 2 circles. The head electronics is 6 hybrid modularization circuit in series. The AWCC
can offer the counts of all detectors of two circles or offer the counts of inside circle and outside circle
respectively.
The AWCC can work on active mode and passive mode. For getting the optimum detection
efficiency，the plugs has different the material and height in each mode.
The AWCC has been tested for a long time. At passive mode，the device has 30.5% neutron detection
efficiency for measuring 252Cf. Some experiments with the AWCC were performed，such as the measuring
effect about some polythene material in nuclear material，simulative U-Al alloy measurement and the
measurement of some new work samples with high enrichment uranium，and some primary results were
obtained.
7 Burnup Estimated by Isotopic Distribution of Xe
Released From Reprocessing Plant
ZHANG Xiao-zhi，LI An-li，ZHANG Ji-long，WANG Chen，WANG Lin-bo
Isotopic distribution of fission-produced Xe changes with burnup of spent fuel.
240Pu/239Pu
atom
number ratio depends on burnup，which is a criterion to distinguish civilian Pu from military Pu. This is a
key of nuclear safeguards at reprocessing plant.
Suitable Xe isotopes are selected in order to estimate burnup with measured Xe isotopic ratios. The
atom number ratios of
131Xe/134Xe
and132Xe /134Xe were calculated by Origen 2.1 code with
235U
abundance of 3.0% in fuel elements for several types of reactor. The calculated result shows that
131Xe/134Xe
and132Xe /
134Xe
atom number ratios are obviously correlated with burnup of the spent fuel
element.
A simulative experimental measurement was carried out. 4 powder samples in quartz tubes (14 mg
U3O8，3.0% 235U) were put at the core region of heavy water research reactor at CIAE for 47 and 81 days，
respectively. The generated gas in the samples was diluted by air to simulate the real gas in stack. The
isotopic ratios of Xe were measured by noble gas mass spectrometer.
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The abundance of
129
Xe is rather high in atmosphere，but is negligible in fission gas. Thus，129Xe
would be a criterion in background subtractions for other isotopes. The atom number ratios of 131Xe/134Xe
and
132
Xe/134Xe，inferred after background subtraction，were compared with the calculated ones.
Experimental result shows that the burnup estimated from atom number ratios of
131
Xe/134Xe and
132
Xe/134Xe with 1.7% measurement precision is indicated with 2.3 GW·d/t. It is easy to determine
whether the spent fuel element is at high or low burnup range.
The fission-producing noble gas is easily released into environment during reprocessing and not
accumulated in the region of concerned facilities. Hence，the ratios of noble gas in stack just give enough
information for verifying independently what is being reprocessed but not divulging other sensitive
information. Moreover，the gas sampling at stack is far away from the center of nuclear facilities，and it
doesn’t interfere the reprocessing plant’s running plan.
It is concluded that measurements of isotopic distribution of Xe released from stack gas at
reprocessing plants could be an effective and low-intrusive method for safeguards inspection.
8 Determination of Plutonium Isotope Ratio of PuO2 Particles
by Secondary Ion Mass Spectrometry
WANG Lin-bo，LI Jing，ZHAO Yong-gang，LI Jing-huai，CHANG Zhi-yuan
The isotope ratio of plutonium in plutonium-bearing particles is essential information for identifying
its origin. This is meaningful for nuclear safeguards and nuclear forensic. Plutonium isotope ratio
239Pu/240Pu in plutonium dioxide particle was determined precisely by secondary ion mass spectrometry
(SIMS) in this study.
Sample preparation procedure was simulated with CeO2 particles. Sputtering a thin layer of carbon is
considered the best method for fixing the plutonium particle on the surface of the graphite sample disk
with the lowest interference of matrix. The prepared plutonium particle was scanned in scanning electron
microscope and located，the sample disk was introduced into SIMS sample chamber，plutonium particles
were relocated and determined with microscope mode and probe mode.