采用快慢中子符合的中子探测器

采用快慢中子符合的中子探测器,可用于强γ射线本底下探测中子,并应用于冷聚变实验中。使用~(252)Cf中子放射源时,探测效率为3×10~(-3),对慢中子峰的FWHM分辨为25%。     

含硼塑料闪烁体的中子-γ甄别

含硼塑料闪烁体在中子探测方面具有许多优良特性,由于~(10)B具有很高的慢中子俘获截面,氢具有很高的快、慢中子散射截面,因此能够实现对慢中子和快中子的同时探测。为有效区分中子和γ事件,用过零时间法对含硼塑料闪烁体BC454的中子-γ甄别能力进行了实验测量。三组对比实验的结果表明,脉冲上升时间谱存在三个明显的高斯分布,它们从左到右依次分别对应于γ射线、快中子和慢中子,因此BC454具有一定的中子-γ甄别能力,能够进行快、慢中子的同时测量。     

~6Li玻璃闪烁体中子脉冲的等效电子能量

ST-602型铈(Ce)激话的~6Li玻璃闪烁体是探测慢中子、特别是热中子的高效率探测器。3mm厚的这种闪烁体对热中子的探测效率可达90％以上,但它对γ射线的探测效率也较高,这对于中子探测来说是极为不利的。因为中子束经常伴随γ射线。为了合理使用~6Li玻璃闪烁体探测中子并有效地甄别掉γ射线,为了正确地选择γ射线的能量,以便使~6Li玻璃探测器能同时用作γ射线和中子的探测元件,因此研究中子、电子(或     

不同材料中子反射与屏蔽效应研究

利用Monte Carlo粒子输运计算程序Super MC对厚度1-5 cm的多种材料进行中子反射和屏蔽性能分析计算。这些材料包括金属材料铍、铅、铜、含硼钢以及~(238)U和非金属材料聚乙烯、氢化锂、混凝土,中子能段选取10~(-5) e V-20 MeV。结果显示,中子反射能力和屏蔽性能都会随着材料厚度而增加,但增加的幅度逐渐减小。铍和聚乙烯在中子反射和屏蔽方面性能优越,而常用来屏蔽γ射线的铅在这两方面性能都是8种材料中最差的。~(238)U只在材料厚度很小时性能卓著,随着材料厚度增加,其性能便远不如大部分材料。考虑到聚乙烯的力学性能较差,在屏蔽材料的选择上有很大的限制,所以在8种材料中,铍的综合性能相对较好。     

海底原位PGNAA探测含水率对中子分布影响研究

运用MC法模拟14 MeV快中子进入沉积物后的物理过程,得到了不同深度下沉积物中子能谱分布,分析了0~1 eV中子在沉积物中横、纵向分布规律以及含水率与Cl~-浓度对中子分布的影响,探讨了模拟条件下0~1 eV中子扩散的最大深度范围。结果表明:沉积物含水率对中子能谱分布和0~1 eV中子横、纵向分布均有显著影响,进行中子活化的最佳深度为2 cm处;0~1 eV中子扩散的最大深度范围为20~40 cm;海水中Cl~-浓度对中子吸收有影响但影响不显著。     

电晕放电慢中子计数管

探测慢中子常用的三氟化硼(BF_3)正比计数管,其主要缺点是:脉冲幅度较低,应用温度范围较窄,抗γ本底辐射能力低(γ辐射最大本底一般不超过100R/hr)。自L.Colli与U.Facchini在1952年指出可以在电晕放电的本底上记录强电离粒子以来,数种电晕放电慢中子计数管已投入了应用。我厂也研制成功J701Z型电晕放电慢中子计数管(简称“电晕管”)。电晕管具有脉冲幅度高(可达数百mV),坪区范围宽(大于2kV),寿命长(几乎无限),热稳定性高(玻壳管可达200℃,金属陶瓷管可达300℃),在强γ本底辐射下工作稳定等优点。可应用于反应堆工程、石油勘探中的中子测井及中子测水分等慢中子测量仪器中。     

硼聚乙烯与铅硼聚乙烯屏蔽性能测试与模拟分析

对屏蔽材料硼聚乙烯、铅硼聚乙烯进行了γ射线、241 Am-Be源中子的屏蔽性能测试,再用蒙卡软件MCNP对材料的屏蔽测试过程进行模拟分析.结果表明模拟结果与实测值相符很好,实测铅硼聚乙烯对60Co与137Cs的γ射线线衰减系数分别为0.212 cm-1和0.381 cm-1,模拟的线衰减系数为0.209 cm-1和0.370 cm-1;硼聚乙烯对241Am-Be源中子的宏观分出截面计算值与实测值分别为0.197 cm-1和0.193 cm-1,铅硼聚乙烯计算值与实测值分别为0.181 cm-1和0.173 cm-1,说明铅硼聚乙烯对中子和γ射线均具有很好的屏蔽效果,用MCNP软件可模拟屏蔽材料对中子和γ射线的屏蔽性能.     

中子与γ射线辐照对B4C/环氧树脂屏蔽材料性能的影响

中子与γ射线辐照对屏蔽材料性能的影响直接关系到了核设施的运行安全性。本研究以B4C/环氧树脂屏蔽材料作为研究对象,对比了在1 MGy γ射线及叠加1.19×1015cm-2中子辐照两种辐照环境下屏蔽材料力学性能、断口组织形貌、特征化学产物及热稳定性能的变化规律。结果表明：持续约11.6 d的γ射线辐照及叠加持续约3 h的中子辐照后屏蔽材料力学性能持续降低,但均未降低到辐照前的50%以下,屏蔽材料在此条件下产生了辐照降解,但未发生失效。与单独的γ射线辐照相比,叠加中子辐照后屏蔽材料1H-NMR图谱δ=7附近峰的强度没有明显变化,说明未继续发生苯环上C-H键的断裂。屏蔽材料热失重50%质量损失温度T50%由辐照前的526.3℃降低到了γ射线辐照后的453.2℃及γ射线叠加中子辐照后的463.9℃,屏蔽材料辐照后热稳定性降低。     

MC法优化设计~(252)Cf中子源辐射屏蔽装置

利用MCNP5程序构建了屏蔽装置模型,并模拟了聚乙烯、含质量分数10%硼的石蜡、重水、石墨和铅等材料的中子慢化和屏蔽效果,以及铁对γ射线的屏蔽效果。当中子慢化剂聚乙烯的厚度达5 cm时,透过慢化层发射出的中子注量率达到最大值为5.40×10-4m~(-2)s~(-1)。中子屏蔽层含硼石蜡厚度为33 cm并且γ屏蔽层铁厚度为4 cm时,由中子和γ射线产生的年有效剂量之和满足国家标准相关限值要求。     

液体闪烁探测器中子γ射线测量研究

液体闪烁体对中子的响应函数,利用O5S蒙特卡罗程序进行计算,γ射线响应函数利用MATHA蒙特卡罗程序进行计算。分别对液体闪烁探测器(n,γ)分辨品质测量、Am-Be中子源的中子谱和γ射线能谱的测量、铁球在D-T中子照射下泄漏中子和伴生γ射线能谱的测量、水球在D-T中子照射下泄漏中子和伴生γ射线能谱的测量等内容进行了介绍,并对结果进行了分析和讨论。     

Chemical interaction between granular B4C and 304L-type stainless steel materials used in BWRs in Japan

Chemical reactions between stainless steel and boron carbide were investigated using the materials applied for control rods in BWRs in Japan, specifically 304L-type stainless steel and granular boron carbide. The reaction region consisted of 2–4 layers, in which the significant composition variation of each element was detected, especially for B and C. Assuming that the reaction layer growth obeys the parabolic law, the effective rate constant between 304L-type stainless steel and granular boron carbide was evaluated to be approximately one order of magnitude smaller than the previously reported values for boron carbide pellets or powers. This difference might originate from the loose contact between the stainless steel and the granular boron carbide in the present study. Regarding liquefaction progress, the stainless steel components were selectively dissolved in the melt; consequently, the unreacted boron carbide tended to remain.     

浸酯涂硼——中子探测器灵敏层制作方法

介绍了两种价廉、实用的中子探测器浸酯涂硼方法--浸没涂硼和手工刷涂,并给出涂抹细节。对不锈钢片的多次实验表明：浸没涂硼时脂硼比λ的最小值为5.0;手工刷涂时λ的最佳值为0.2,最佳方法是先在待涂面上滴混合液,再撒上硼粉刷涂。采用该方法研制一圆柱形硼衬正比计数管,用标准Am-Be中子源辐照测得了其输出脉冲;用不同厚度聚乙烯管包裹后测得了Am-Be源的中子谱。对中子谱参数的分析表明,两种硼膜制作过程简单、方法可行,为业内人员提供了很有意义的参考。     

The effect of dose rate on the response of austenitic stainless steels to neutron radiation

Depending on reactor design and component location, austenitic stainless steels may experience significantly different irradiation dose rates in the same reactor. Understanding the effect of dose rate on radiation performance is important to predicting component lifetime. This study examined the effect of dose rate on swelling, grain boundary segregation, and tensile properties in austenitic stainless steels through the examination of components retrieved from the Experimental Breeder Reactor-II (EBR-II) following its shutdown. Annealed 304 stainless steel, stress-relieved 304 stainless steel, 12% cold-worked 316 stainless steel, and 20% cold-worked 316 stainless steel were irradiated over a dose range of 1–56 dpa at temperatures from 371 to 440 °C and dose rates from 0.5 to 5.8 × 10⁻⁷ dpa/s. Density and tensile properties were measured for 304 and 316 stainless steel. Changes in grain boundary composition were examined for 304 stainless steel. Swelling appears to increase at lower dose rates in both 304 and 316 stainless steel, although the effect was not always statistically significant. Grain boundary segregation also appears to increase at lower dose rate in 304 stainless steel. For the range of dose rates examined, no measurable dose rate effect on tensile properties was noted for any of the steels.     

压水堆条件下锌对奥氏体不锈钢腐蚀性能的影响

模拟压水堆一回路冷却剂环境,对316和304奥氏体不锈钢在不加锌和加锌浓度为50ppb的315℃溶液中进行了两组500h腐蚀实验。结果表明,锌能有效地降低两种材料的均匀腐蚀速率,加锌后表面氧化膜厚度变薄,氧化膜形貌和成分也有明显改变,304不锈钢表面有大量针状腐蚀产物出现。     

304奥氏体不锈钢在硝酸溶液体系中的电化学腐蚀行为

采用极化曲线测量法对304不锈钢在硝酸溶液体系中的电化学耐蚀性能进行了测试,分别研究了在硝酸溶液中添加硝酸盐、草酸、乙酸、柠檬酸等成分对304不锈钢电化学腐蚀行为的影响。结果表明,在硝酸溶液中,硝酸盐的加入能够抑制不锈钢的电化学腐蚀,而草酸能够显著增强溶液对不锈钢的电化学腐蚀能力,在硝酸和草酸溶液体系中加入1g/L柠檬酸后,自腐蚀电流由6.02μA/cm~2上升到22.8μA/cm~2,对电流腐蚀有较明显的促进作用。     

HFETR堆芯不锈钢反射层的研究

用快中子分出截面法估算了不锈钢棒对压力容器处快中子注量率的减弱程度,用二维多群粒子输运程序ＤＯＴ３．５计算了不锈钢棒的γ释热,计算了高通量工程试验堆（ＨＦＥＴＲ）设计工况和目前运行工况下不锈钢壁面温度和中心温度,对承重栅板的安全性进行了讨论,分析了不锈钢棒对堆芯的性能影响。结果表明,用不锈钢代替铝棒作最外层反射层,堆芯安全性能不变,压力容器处快中子注量率被大幅度降低。     

304奥氏体不锈钢在不同浓度硝酸中的电化学行为

采用动电位极化曲线测量、开路电位测量等技术,研究了304奥氏体不锈钢在不同浓度硝酸溶液中的电化学腐蚀行为,并对304奥氏体不锈钢在硝酸溶液中的电化学反应历程进行了探讨。结果表明：304奥氏体不锈钢在硝酸溶液中具备不锈钢典型的极化曲线特征,有多个钝化区和过钝化区;硝酸浓度升高促进不锈钢表面钝化膜的生成,使开路电位向正电位方向移动,降低了硝酸溶液对不锈钢的腐蚀倾向,同时,随着硝酸浓度的升高,不锈钢的点蚀电位升高,提高了不锈钢耐点蚀能力;在硝酸溶液中,不锈钢的腐蚀速率同时受到酸度和硝酸根浓度的影响,二者相互矛盾,导致硝酸浓度对腐蚀速率的影响呈不规律性。结果表明,在0.5 mol/L硝酸中,不锈钢的腐蚀速率最高。     

模拟二回路环境下乙醇胺与3-甲氧基丙胺复合水工况对304L不锈钢的缓蚀特性

乙醇胺（ETA）作为碱化剂对减缓核电站二回路系统的腐蚀有良好效果。为增强ETA的缓蚀效果,进一步减少腐蚀产物向蒸汽发生器的迁移,本文研究了3 mg/L ETA与不同浓度（01、03、05、08 mg/L）3 甲氧基丙胺（MPA）复合水工况对304L不锈钢的缓蚀特性,通过高压釜高温挂片模拟二回路条件,并通过微分电容曲线与原子力显微镜（AFM）力曲线测试结合的表征手段对其缓蚀机理进行了探讨,并对其进行了形貌表征。微分电容曲线表明,ETA+MPA复合水工况可更好地抑制304L不锈钢表面的腐蚀,溶液中有机胺分子吸附在金属表面,改变了金属表面的双电层结构,并使金属电极的零电荷电位正移,吸附数据拟合结果符合Flory Huggins等温线模型。AFM力曲线结果显示,有机胺在304L不锈钢表面具有黏附力特征,这验证了不锈钢表面微分电容的变化,即有机胺浓度越大、有机胺的吸附层越致密均匀,在不锈钢表面的覆盖度越大,缓蚀效率越高;高温挂片的形貌分析（SEM）、成分分析（EDS）结果与上述微分电容曲线、AFM力曲线分析结果表现出一致性。     

The effects of low dose rate irradiation and thermal aging on reactor structural alloys

As part of the EBR-II reactor materials surveillance program, test samples of fifteen different alloys were placed into EBR-II in 1965. The surveillance (SURV) program was intended to determine property changes in reactor structural materials caused by irradiation and thermal aging. In this work, the effect of low dose rate (approximately 2 × 10⁻⁸ dpa/s) irradiation at 380–410°C and long term thermal aging at 371°C on the properties of 20% cold worked 304 stainless steel, 420 stainless steel, Inconel X750, 304/308 stainless weld material, and 17-4 PH steel are evaluated. Doses of up to 6.8 dpa and thermal aging to 2994 days did not significantly affect the density of these alloys. The strength of 304 SS, X750, 17-4 PH, and 304/308 weld material increased with irradiation. In contrast, the strength of 420 stainless steel decreased with irradiation. Irradiation decreased the impact energy in both Inconel X750 and 17-4 PH steel. Thermal aging decreased the impact energy in 17-4 PH steel and increased the impact energy in Inconel X750. Tensile property comparisons of 304 SURV samples with 304 samples irradiated in EBR-II at a higher dose rate show that the higher dose rate samples had greater increases in strength and greater losses in ductility.     

In-reactor deformation of solution annealed type 304L stainless steel

Over the past six years at EBR-II, a great deal of information has been obtained on the in-reactor behaviour of solution annealed-Type 304L stainless steel. This information consists of the following: (1) Irradiation induced swelling results in the form of immersion density and transmission electron microscope (TEM) measurements on unstressed material that extends over a temperature range of 395° to 530°C and a neutron fluence range of 1.8 to 9.3 × 10²² n/cm² (E > 0.1 MeV). (2) Irradiation induced creep results from helium pressurized capsules irradiated at a temperature of 415°C. The hoop stress range covered in the experiment was 0 to 27.3 ksi, and the peak neutron fluence obtained to date is 7 × 10²² n/cm² (E > 0.1 MeV). (3) Residual stress measurements (slit tube technique) with complementary TEM gradient studies on stressed and unstressed capsules. (4) Comparative swelling studies of stressed cladding material and unstressed capsule material from encapsulated EBR-II driver fuel experiments over wide ranges of temperature and neutron fluences. The deformation information derived from the four above studies represent an extensive data base from which to obtain an understanding of the in-reactor deformation of austenitic stainless steel. It is the purpose of this paper to review our information on the in-reactor deformation of solution annealed Type 304L stainless steel.