活度微量热计研究进展

为了满足热功率为10μW水平的仅和β放射源活度校准需要,研制了1台活度微量热计测量装置。该装置采用半导体测温元件组合（Bi2Te3＋Sb2Te3p型与Bi2Te3＋Bi2Se3n型）作为测温元件,采用等温双杯量热计的技术原理和使用计算机软件根据测温信号的极性用电功率在空白测量杯上进行自动补偿确定放射源的热功率。     

活度微量热计测量氚样品应用

为了满足热功率为10μW水平的α和β放射源活度校准需要,研制了1台等温双杯活度微量热计测量装置。该装置采用半导体测温元件组合（Bi2Te3＋Sb2Te3p型与Bi2Te3＋Bi2Se3n型）作为测温元件、采用等温双杯量热计的技术原理和使用计算机软件根据测温信号的极性用电功率在空白测量杯上进行自动补偿确定放射源的热功率。     

微量热计测量低活度氚

针对国际热核聚变实验堆氚增殖系统(ITER-TBS)中微量氚活度的绝对测量需求,设计研制了微量热计。采用固体含氚样品对微量热计的输出热电势(Eout)-输入热功率(P)进行标定,Eout-P关系式具有很好的线性关系,灵敏度系数达到0.13V/W。研制的微量热计最低检测限低于0.2μW,量热杯体积大于500mL,该性能指标预期可定量测量ITER-TBS系统中各种复杂形态的微量氚活度,具有很好的应用前景。     

低能β核素微热量热计研制

为对小体积包容体内β核素的放射性活度进行非破坏性测量,研制了低能β核素微热量热计。本工作描述了该量热计工作原理及装置组成,并对装置进行了性能测试和实验验证,对热功率测量结果进行了不确定度评定。输入功率为500μW时,测量结果的不确定度为0.96％（k=2）;输入功率为48μW时,测量结果与输入功率标准值间的偏差为4.2%。     

医用同位素微量热计性能研究

为实现医用同位素核纯度和放射性活度的准确测量,本工作根据量热法测量原理设计一套微量热装置,包括量热系统、温控系统以及测控软件,并对装置的性能进行测试和实验验证。结果表明,同位素量热计的基线电势稳定,复现性良好。最低探测限可至3.1 μW,能够实现医用同位素放射性活度测量的全覆盖。在10 μW~1 mW的测量范围内,当输入热功率<100 μW时,单次测量结果与输入热功率标准值之间的最大偏差<1.5%;当输入功率>100 μW时,测量结果和输入热功率偏差<0.5%。利用该装置成功实现了医用同位素¹⁴C样品的活度测量,测量结果与标准值在不确定度范围内相吻合,不产生放射性废物和化学废物,不破坏样品,操作简单方便。后期可通过研究测量预热和保温装置,进一步提高装置的适应能力和探测下限。     

电子束吸收剂量量热计

研制了一种用于电子束吸收剂量深度分布和电子束能量测量的多层石墨量热计及一种用于吸收剂量测量和薄膜剂量计校准的石墨量热计校准装置,并介绍了两种量热计的结构和在中国原子能科学研究院14MeV加速器上的研究结果。     

裂变热电偶时间响应特性研究

裂变热电偶是一种能在强γ辐射环境下测量中子的快响应小型化探测器。本文从理论上估算了裂变热电偶的时间响应,好于1μs,并通过热阶跃响应实验验证了理论计算,     

核电站反应堆热电偶机械密封件研究

压水反应堆堆内温度测量装置是监测反应堆热功率输出及堆内功率分布情况、保证反应堆安全运行的手段之一。该装置以热电偶作为测量元件,用不锈钢作保护外套。由于热电偶有一定的使用寿命,须定期更换,为此研制了一种适用于核电厂反应堆堆内温度测量的热电偶密封结构。这种结构既能在高温高压下密封牢靠,没有泄漏和渗漏,又能在带放射性环境下方便地逐根更换热电偶。它是一种结构简单实用的新型密封结构。     

核电站反应堆热电偶机械密封件研究

压水反应堆堆内温度测量装置是监测反应堆热功率输出及堆内功率分布情况、保证反应堆安全运行的手段之一。该装置以热电偶作为测量元件,用不锈钢作保护外套。由于热电偶有一定的使用寿命,须定期更换,为此研制了一种适用于核电厂反应堆堆内温度测量的热电偶密封结构。这种结构既能在高温高压下密封牢靠,没有泄漏和渗漏,又能在带放射性环境下方便地逐根更换热电偶。它是一种结构简单实用的新型密封结构。     

压水堆堆芯新型非能动热声测温装置

本文基于热声效应原理设计出一种新型的测温装置,用于弥补现今核反应堆内以热电偶等方式测温缺乏非能动特性的不足,可以提高在严苛环境下仪表测量的可靠性和安全性本文利用DeltaEC热声计算软件对设计的热声测温装置各个组件尺寸进行了优化,目标是使得各个组件组成的系统性能最佳,即在同等工作条件下装置内气体震荡幅度最高。经过优化后,通过改变热端温度找到了热端温度与装置内声波频率的对应关系,近似为一条一次函数直线,因此新型的测温装置可以有效地实现非能动测量。     

CPR1000核电厂反应堆功率标定系统设计及验证

简要介绍CPR1000核电厂反应堆功率标定系统的运算原理、系统结构、验证过程及结果。采用IAPWS-IF97水和蒸汽物性模型、焓差-热功率原理计算核电厂核蒸汽供应系统热功率与反应堆核功率。该套系统于岭澳核电站3号机组调试启动中成功完成了反应堆功率测量系统、控制系统的实态标定。实际应用表明功率标定系统能够可靠地完成CPR1000核电厂堆芯功率标定。     

Thermal power calibrations of the IPR-R1 TRIGA reactor by the calorimetric and the heat balance methods

Since the first nuclear reactor was built, a number of methodological variations have been evolved for the calibration of the reactor thermal power. Power monitoring of reactors is done by means of neutronic instruments, but its calibration is always done by thermal procedures. The purpose of this paper is to present the results of the thermal power calibration carried out on March 5th, 2009 in the IPR-R TRIGA reactor. It was used two procedures: the calorimetric and heat balance methods. The calorimetric procedure was done with the reactor operating at a constant power, with primary cooling system switched off. The rate of temperature rise of the water was recorded. The reactor power is calculated as a function of the temperature-rise rate and the system heat capacity constant. The heat balance procedure consists in the steady-state energy balance of the primary cooling loop of the reactor. For this balance, the inlet and outlet temperatures and the water flow in the primary cooling loop were measured. The heat transferred through the primary loop was added to the heat leakage from the reactor pool. The calorimetric method calibration presented a large uncertainty. The main source of error was the determination of the heat content of the system, due to a large uncertainty in the volume of the water in the system and a lack of homogenization of the water temperature. The heat balance calibration in the primary loop is the standard procedure for calibrating the power of the IPR-R1 TRIGA nuclear reactor.     

Thermal power calibrations of the IPR-R1 TRIGA reactor by the calorimetric and the heat balance methods

Since the first nuclear reactor was built, a number of methodological variations have been evolved for the calibration of the reactor thermal power. Power monitoring of reactors is done by means of neutronic instruments, but its calibration is always done by thermal procedures. The purpose of this paper is to present the results of the thermal power calibration carried out on March 5th, 2009 in the IPR-R TRIGA reactor. It was used two procedures: the calorimetric and heat balance methods. The calorimetric procedure was done with the reactor operating at a constant power, with primary cooling system switched off. The rate of temperature rise of the water was recorded. The reactor power is calculated as a function of the temperature-rise rate and the system heat capacity constant. The heat balance procedure consists in the steady-state energy balance of the primary cooling loop of the reactor. For this balance, the inlet and outlet temperatures and the water flow in the primary cooling loop were measured. The heat transferred through the primary loop was added to the heat leakage from the reactor pool. The calorimetric method calibration presented a large uncertainty. The main source of error was the determination of the heat content of the system, due to a large uncertainty in the volume of the water in the system and a lack of homogenization of the water temperature. The heat balance calibration in the primary loop is the standard procedure for calibrating the power of the IPR-R1 TRIGA nuclear reactor.     

Development of Dielectric Microcalorimeter

A prototype dielectric microcalorimeter was fabricated from 0.99SrTiO₃-0.01SrTa₂O₆, a dielectric material. To estimate the operation performance of the microcalorimeter, a simple evaluation model is introduced by employing the block diagram formalism. Pulse height and fall time constant of the detection signal are analyzed as a function of the thermal conductance of the heat link. The dielectric microcalorimeter maintained at 100mK was irradiated by 5.5 MeV α-particles emitting from an ²⁴¹Am source. Responsivity was found to be 2.7x10^?19C/eV.     

Study of a 105 GHz short-pulse dummy load for the electron cyclotron resonance heating system with the quasi-optical method

The accurate power measurement is important for an ECRH system in tokamak. The dummy load is designed and developed for the measurement of the millimeter wave power. This work analyzes the dummy load based on the quasi-optical method and the ray tracing method. The reflectivity and thermal deposition of the dummy load have been considered to ensure the safety of the entire system. High-power tests have been carried out at a 105 GHz/500 kW ECRH system. The results of the tests indicate that the designed dummy load is stable and valid.     

直流电流传感器高精度校准系统设计

直流电流传感器（DCCT）的测量精度是高精度直流稳流电源输出电流精度的关键影响因素之一。针对DCCT优于1×10^-5的电流测量精度,本文提出了一种通过用高准确度的直流比较仪（DCC）测量电阻原理设计的校准系统,可实现DCCT 0~400 A测量范围内的高精度校准,且在整个校准范围内系统测量的相对不确定度优于1.1 ppm。通过实验测试DCCT的准确度和线性度验证了整个校准系统的精度,为进一步提高磁铁电源的电流精度提供了保障。     

医院中子照射器物理启动

医院中子照射器是专门用于硼中子俘获治疗的核装置。在堆芯相对两侧,设有热中子束流和超热中子束流用于治疗,另外,在热中子束流内引出1条热中子束流用于病人血硼浓度测量。本文介绍其物理启动的6个实验,实验结果表明：满功率最大运行时间为12 h,最终后备反应性为4.2 mk,满功率运行时各工艺房间辐射水平满足设计辐射分区要求,4.2 mk反应性释放实验证明医院中子照射器具有固有安全特性。     

高灵敏中子剂量当量率仪的能量和剂量响应校准

参考国内标准化和国际标准化组织（ISO）及国际电工委员会（IEC）对中子（率）仪校准的相应标准,对已设计、加工、组装完毕的1台长圆柱状、较大体积、高灵敏中子剂量当量率仪,在宽能区（热中子～20MeV）内进行了能量响应和剂量灵敏度的实验校准。本文介绍了仪器能谱响应函数曲线实验校准方法、实验方案和结果分析,给出了仪器的综合灵敏度48.9s－1/（µSv•h－1）及在3种特征能谱中的剂量响应。     

Indication Lowering of Average Magnitude Type Campbelling System in Low Pulse Rate Region

In the nuclear power calibration process of JOYO, the first experimental fast breeder reactor in Japan, the indication of the Intermediate Range Monitoring System (IRMS), employing Average Magnitude type Campbelling System (AMCS), was calibrated with the reactor power at 45.82 kW. The reactor power was then decreased and the nonlinear indication lowering of the IRMS was observed.

In this paper, we present a derivation of an equation representing the AMCS characteristic and show that the indication lowering occurs when the product of mean arrival rate of pulses and their width is small. The computed values based on the derived equation agreed very well with the observed ones in the JOYO IRMS below 45.82 kW, and it was confirmed that the evaluation method was applicable. Furthermore, it was evident from the evaluation that the indication lowering of the JOYO IRMS above 45.82 kW was negligibly small, and did not affect the reactor power ascension testing. Also, it was proved in a following thermal calibration test above 50 kW that the indication lowering was allowably small indeed in this power region.     

The design of “multiple energies from one source” in D-T neutron tube

To obtain multiple monoenergetic neutron sources and realize the on-site calibration of radiation monitoring equipment for nuclear-involved places,the structural characteristics and neutron source features of D-T neutron tube were analyzed;Monte Carlo method was adopted to simulate the effect of interaction between typical materials and different energy neutrons;multilayered shielding materials were combined and optimized to acquire the optimal scheme to shield the neutron sources from the neutron tube.On the base,a tapered alignment filtration construction was designed and Monte Carlo method was employed to simulate the effect of alignment construction.The result showed that the tapered alignment filtration construction can create monoenergetic neutrons including14.1 MeV,0.18 MeV and thermal neutrons and demonstrated good monochrome performance which provides multiple monoenergetic sources for the on-site calibration.