An Epithermal Neutron Source for Cancer Therapy

描述(申请人提供)：我们建议开发一种基于硼中子俘获疗法(BNCT)的商业中子源，用于治疗脑癌和肝癌。含硼化合物被注射到患者的血液中，并在恶性肿瘤中积聚。当受到中子照射时，硼原子优先捕获中子并衰变，释放短程辐射，杀死肿瘤，但不会对其他组织造成明显损害。BNCT主要是使用核反应堆进行测试，这些反应堆又大又贵，不能在临床环境中使用。拟议的污染源相对便宜，足够紧凑，可以在医院使用，并保证安全运行。此外，它将被设计成产生理想的超热中子谱(与反应堆不同)，预计将产生足够短时间处理的中子通量。建议的源包括一个使用D-T反应的射频等离子体快中子源和一个优化的、紧凑的慢化器来减缓和准直中子。这些组分以一种新颖的方式结合在一起，为BNCT产生高通量、最佳的超热中子谱。氚的储存、补充和回收将通过一个独立的真空系统实现，以将辐射和爆炸危险降至最低。我们的设计是新颖的，因为它将使用优化的快中子靶的等离子体中子源与新的慢化剂材料和几何结构相结合，以实现更高的效率和紧凑性。初步分析表明，优化靶阳极的几何形状可以减少与适当设计的慢化剂集成时对发电机的所需功率。这种蒙特卡罗N粒子(MCNP)分析将得到改进，以确定D-T发生器的几何形状，其慢化器配置将产生最佳中子能量的中子束并最大限度地减少功耗。为了便于水冷却，靶阳极上的离子束功率密度将低于每平方厘米150瓦。我们将使用放射治疗应用程序模拟环境(SERA)治疗程序创建用于剂量计算的源模型。为了进行比较，我们将使用MCNP程序计算患者模型中的剂量。快中子发生器将使用研究所开发的计算机代码进行设计。以前的快中子源已经被制造和测试，证明了这些程序的有效性和研究所的技术技能。

DESCRIPTION (provided by applicant): We propose to develop a commercial neutron source for the treatment of brain and liver cancer based on boron neutron capture therapy (BNCT). Boron-containing compounds are injected into a patient's bloodstream and accumulate in a malignant tumor. When irradiated with neutrons, the boron atoms preferentially capture neutrons and decay, releasing short-range radiation that kills the tumor but does not appreciably damage other tissues. BNCT has been tested using primarily nuclear reactors that are large and expensive and cannot be used in a clinical setting. The proposed source is relatively inexpensive, compact enough to be used in hospitals and promises safe operation. In addition, it will be designed to produce an ideal epithermal neutron spectrum (unlike reactors) and is expected to produce enough neutron flux for short treatment times. The proposed source consists of an rf-plasma fast-neutron source using the D-T reaction and an optimized, compact moderator to slow and collimate the neutrons. These components are brought together in a novel fashion to produce a high-flux, optimum epithermal neutron spectrum for BNCT. Tritium storage, replenishing and recycling will be achieved by a self-contained vacuum system to minimize radiation and explosion hazards. Our design is novel in that it integrates the plasma neutron source using an optimized fast-neutron target with new moderator materials and geometries for greater efficiency and compactness. Preliminary analysis shows that optimizing the target-anode's geometry can decrease the required power to the generator when integrated with a properly designed moderator. This Monte-Carlo N-particle (MCNP) analysis will be improved to determine the D-T generator's geometry with a moderator configuration that will produce a neutron beam of optimum neutron energy and minimize power consumption. To facilitate water cooling, the ion beam power density on the target anode will be less than 150 Watts per square cm. We will create a source model for dose calculations using the Simulation Environment for Radiotherapy Application (SERA) treatment program. For comparison, we will calculate the dose in a patient model with the MCNP program. The fast-neutron generator will be designed using computer codes developed at the research institute. Prior fast neutron sources have been fabricated and tested demonstrating the efficacy of these codes and the research institute's technological skill.