Heavy Water Moderator Research Articles

Study of moderator thickness for an accelerator-based neutron irradiation facility for boron neutron capture therapy using the 7Li(p, n) reaction near threshold

Accelerator neutron sources for epithermal neutron capture therapy utilizing the 7Li(p, n) nuclear reaction will require a moderator even in the threshold range of 1.89 to 1.95 MeV. The corresponding neutron energies allow for a thinner reflector and moderator, with less reduction of the epithermal flux. To estimate the useful neutron flux within the epithermal range (4 eV-40 keV), the optimal thickness of a heavy water moderator was determined using the two-dimensional neutron transport SN code DORT. Optimized results are compared with the epithermal fluxes reported for the higher proton energy range, and are found to be inferior. Thus, this study supports the 2.5-3.0 MeV proton energy range for accelerator boron neutron capture therapy.

Criticality calculations with MCNP4A using correlated sampling methods

Two methods of calculating reactivity effects due to small changes in reactor configurations have been implemented in MCNP4A. The pseudo-correlated sampling method makes use of the correlation of particle tracks in problems with well localized perturbations. The true correlated sampling method forces a perturbed particle on the track of a reference particle and adjusts its weight for compensation. It works well for problems of small changes in the composition or density of materials and is not restricted to a localized region of the reactor arrangement. Both methods are using the fission matrix approach for calculating the eigenvalue of the multiplying system rather than the source iteration method. These methods of correlated sampling have been applied to problems in the FRM-II research reactor design, to study small reactivity changes due to installations in the moderator pool and to verify reactivity coefficients due to density changes of the cooling water and the heavy water moderator.

Flooding of a large, passive, pressure-tube light water reactor

A reactor concept has been developed which can survive loss of coolant accidents without scram and without replenishing primary coolant inventory, while maintaining safe temperature limits on the fuel and pressure tubes. The proposed concept is a pressure tube type reactor of similar design to CANDU reactors, but differing in three key aspects. First, a solid SiC-coated graphite fuel matrix is used in place of fuel pin bundles to enable the dissipation of decay heat from the fuel in the absence of primary coolant. Second, the heavy water coolant in the pressure tubes is replaced by light water, which also serves as the moderator. Finally, the calandria tank, surrounded by a graphite reflector, contains a low pressure gas instead of heavy water moderator, and this normally-voided calandria is connected to a light water heat sink. The cover gas displaces the light water from the calandria during normal operation, while during loss of coolant or loss of heat sink accidents it allows passive calandria flooding. Calandria flooding also provides redundant and diverse reactor shutdown. This paper describes the thermal hydraulic characteristics of the passively initiated, gravity driven calandria flooding process. Flooding the calandria space with light water is a unique and very important feature of the proposed pressure-tube light water reactor (PTLWR) concept. The flooding of the top row of fuel channels must be accomplished fast enough so that in the total loss of coolant, none of the critical components of the fuel channel, i.e. the pressure tube, the calandria tube, the matrix and the fuel, exceed their design limits. The flooding process has been modeled and shown to be rapid enough to maintain all components within their design limits. Two other considerations are important. The thermal shock experienced by the calandria and pressure tubes has been evaluated and shown to be within acceptable bounds, while simultaneously cooling the graphite fuel matrix. This has been demonstrated for the bounding scenarios of sustained film boiling and rapid quenching to yield nucleate boiling on the calandria tube outer surface. Finally, although complete flooding renders the reactor deeply subcritical, various steam/water densities can be hypothesized to be present during the flooding process which could cause reactivity to increase from the initially voided calandria case. One such hypothesis which leads to the maximum possible density of the steam/water mixture in the still unflooded calandria space is entrainment from the free surface. It is shown that the steam/water mixture density yielding the maximum reactivity peak cannot be achieved by entrainment because it exceeds thermohydraulically attainable densities of steam/water by an order of magnitude.

Characterisation of an Accelerator-Based Neutron Source for BNCT of Explanted Livers

An accelerator-based thermal neutron source for BNCT of the explanted liver was designed using the MCNP code. Neutrons are generated via (d,n) reactions by 7 MeV deuterons bombarding a beryllium target. The therapy constraints were approached by simulating an irradiation cavity placed inside a graphite reflector parallelepiped containing a heavy-water moderator in turn enclosing the beryllium target. The experimental verification was performed at the Laboratori Nazionali di Legnaro (Italy). The thermal and epithermal neutron flux was measured at various positions in the irradiation cavity by means of activation techniques employing bare and cadmium covered indium foils. Further measurements were performed with BF 3 detectors. The fast neutron component of the dose equivalent and the energy spectrum above 100 keV were assessed by means of a recently developed technique employing variable threshold superheated drop detectors. The prompt gamma ray dose was measured with 7 LiF TLDs.'

Large Passive Pressure Tube Light Water Reactor with Voided Calandria

A reactor concept has been developed that can survive loss-of-coolant accidents (LOCAs) without scram and without replenishing primary coolant inventory while maintaining safe temperature limits on the fuel and pressure tube. The proposed concept is a pressure tube reactor of similar design to Canada deuterium uranium reactors but differing in three key aspects. First, a solid silicon carbide-coated graphite fuel matrix is used in place of fuel pin bundles to enable the dissipation of decay heat from the fuel in the absence of primary coolant. Second, the heavy water coolant in the pressure tubes is replaced by light water, which also serves as the moderator. Finally, the calandria tank, surrounded by a graphite reflector, contains a low-pressure gas instead of heavy water moderator, and this normally voided calandria is connected to a light water heat sink. The cover gas displaces the light water from the calandria during normal operation while during a LOCA or loss of heat sink accident, it allows passive calandria flooding. Calandria flooding also provides redundant and diverse reactor shutdown. The fuel elements can operate under post-critical-heat-flux conditions even at full power without exceeding fuel design limits. The heterogeneous arrangement of the fuel and moderator ensures amore » negative void coefficient under all circumstances. Although light water is used as both coolant and moderator, the reactor exhibits a high degree of neutron thermalization and a large prompt neutron lifetime, similar to D{sub 2}O-moderated cores. Moreover, the extremely large neutron migration length results in a strongly coupled core with a flat thermal flux profile and inherent stability against xenon spatial oscillations.« less

Long-Term Performance of Atmospheric-Detritiation Dryers

ABSTRACTTritiated heavy water (DTO) accumulates in the process systems of a CANDU®a reactor due to neutron capture by the heavy-water moderator and coolant. After twelve years of service, the moderator in the Point Lepreau Generating Station has reached approximately 1.6 TBq/kg, and the total inventory exceeds 300 PBq. Point Lepreau uses nine desiccant dryers to control airborne heavy water and tritium. Ranging in size from 1 000 m3/h to 6 800 m3/h, the majority are single-bed, cocurrent-regenerated units filled with 13X or 4A molecular sieve. These dryers have operated almost continuously for twelve years without a significant breakdown. During the last thirteen years, their availability has exceeded 99% and they have routinely dried air to a dew-point temperature of −60°C or below. Tritium emissions from the dried areas in the reactor building remain a small fraction of the tritium released into the reactor building. The keys to the success of this detritiation system are the mechanical simplicity of th...

Neutronics and Engineering Design of the Aqueous-Slurry Accelerator Transmutation of Waste Blanket

A conceptual target and blanket design for an accelerator transmutation of waste system capable of transmuting the high-level waste stream from 2.5 light water reactors is described. Typically, four such target-blanket designs would be served by a single linear accelerator. The target consists of rows of solid tungsten rod bundles, cooled by heavy water and surrounded by a lead annulus. The annular blanket, which surrounds the target, consists of a set of actinide-oxide-slurry-bearing tubes, each 3 m long, surrounded by heavy water moderator. Heat is removed from the slurry tubes by passing the slurry through an external heat exchanger. Long-lived fission products are burned in regions that are separate from the actinides. Using the Monte Carlo codes LAHET and MCNP, a conceptual design for a beam current of 62.5 mA/target of 1.6-GeV protons has been developed. Preliminary engineering analyses on key system components have been performed. A preliminary layout of the concept and the associated primary-heat transport subsystems was developed, demonstrating a multiple-containment-boundary design philosophy.

Parametric Systems Studies of the Aqueous-Based (Slurry) Blanket Concept for Accelerator Transmutation of Waste

Transmutation of long-lived nuclear waste currently stored in spent reactor fuels may represent an attractive alternative to deep geologic disposal. The aqueous-based accelerator transmutation of waste (ATW) concept uses a proton accelerator to produce a 1.6-GeV, 250-mA ( ca. 400 MW) beam that is split four ways and directed to four D{sub 2}O-cooled solid W-Pb composite targets. Each target in turn is centered in a heavy water moderated, highly multiplying, actinide (oxide)-slurry blanket. The target-blanket system for ATW resides at an interface separating two major systems that are crucial to the economic and technical success of the concept: (a) the high-energy (power-intensive) accelerator delivering 0.8 to 1.6 GeV protons to the high-Z spallation neutron source and (b) the chemical-plant equipment (CPE) that provides feedstock appropriate for efficient and effective transmutation. Parametric studies have been performed to assess the effects of the target-blanket on overall system performance with regard to neutron economy, chemical-processing efficiency and thermal-hydraulic design options. Based on these parametric evaluations, an interim base-case aqueous-slurry ATW design was selected for more detailed analysis. This base-case target-blanket consists of an array of Zr-Nb pressure tubes placed in a heavy water moderator surrounding a heavy-water-cooled W-Pb target. Neutronics and thermal-hydraulic calculationsmore » indicate that each of the four ATW target-blanket modules operating with a neutron multiplication k{sub eff} = 0.95 can transmute the actinide waste and the technetium and iodine waste from ca. 2.5 light water reactors. By recovering the fission heat, sufficient electricity can be produced both to operate the accelerator and to supply power to the grid for revenue generation. These broad-based parametric studies have provided guidance to a preliminary conceptual engineering design of the aqueous-slurry ATW blanket concept.« less

Tritium activities in Canada

Canadian tritium activites comprise three major interests: utilites, light manufacturers, and fusion. There are 21 operating CANDU reactors in Canada; 19 with Ontario Hydro and one each with Hydro Quebec and New Brunswick Power. There are two light manufacturers, two primary tritium research facilities (at AECL Chalk River and Ontario Hydro Technologies), and a number of industry and universities involved in design, construction, and general support of the other tritium activities. The largest tritum program is in support of the CANDU reactors, which generate tritium in the heavy water as a by-product of normal operation. Currently, there are about 12 kg of tritium locked up in the heavy water coolant and moderator of these reactors. The fusion work is complementary to the light manufacturing, and is concerned with tritium handling for the ITER program. This included design, development and application of technologies related to Isotope Separation, tritium handling, (tritiated) gas separation, tritium-materials interaction, and plasma fueling.

Multisphere Spectrometry and Analysis of TEPC and REM Meter Results Around a Heavy Water Moderated Reactor

The NRU reactor is a vertically-oriented, heavy water moderated and cooled research reactor. Water and iron shielding above the core permit only a small amount of leakage neutrons through the top deck plate. Two leakage neutron spectra determined using a multisphere, or Bonner sphere, system are reported here. The spectra are used to analyse measurements made with a moderator-type neutron dose equivalent meter (i.e. a rem meter) and a tissue-equivalent proportional counter (TEPC) system. Rem meter readings are not reproduced by folding the response function of the instrument with the fluence spectra. Based on TEPC and Bonner sphere data, ambient dose equivalents are found to be 30 and 17 µSv.h-1 (±25%), in the two locations investigated, for a reactor power level of 107 MWt.

A design study for an accelerator-based epithermal neutron beam for BNCT

An achievable design concept for a boron neutron capture therapy (BNCT) facility, based on a high-current, low-energy proton accelerator, is described. Neutrons are produced within a thick natural lithium target, under bombardment from protons with an initial energy between 2.5 and 3.0 MeV. The proton current will be up to 10 mA. After gamma -ray filtering, the neutrons are partially moderated to epithermal energies within a heavy-water moderator, poisoned with 6Li to remove thermal neutrons. Monte Carlo modelling has been used to predict system performance in terms of neutron fluence rate and neutron and gamma -ray dose at the patient position. The relationship between the system performance and key parameters, such as proton energy, moderator depth and 6Li concentration, has been investigated. With a proton current of 10 mA, the facility is capable of providing a therapy beam with a useful neutron fluence rate of 109 cm-2 and a neutron dose per unit fluence of less than 6*10-13 Gy cm2, with a gamma -ray contamination of the therapy beam of about 10-13 Gy cm2.

Radiocarbon Dispersion around Canadian Nuclear Facilities

Canadian deuterium uranium (CANDU) pressurized heavy-water reactors produce 14C by neutron activation of trace quantities of nitrogen in annular gas and reactor components (14N(n,p)14C), and from 17O in the heavy water moderator by (17O(n,α)14C). The radiocarbon produced in the moderator is removed on ion exchange resins incorporated in the water purification systems; however, a much smaller gaseous portion is vented from reactor stacks at activity levels considerably below 1% of permissible derived emission limits. Early measurements of the carbon speciation indicated that >90% of the 14C emitted was in the form of CO2. We conducted surveys of the atmospheric dispersion of 14CO2 at the Chalk River Laboratories and at the Pickering Nuclear Generating Station. We analyzed air, vegetation, soils and tree rings to add to the historical record of 14C emissions at these sites, and to gain an understanding of the relative importance of the various carbon pools that act as sources/sinks within the total 14C budget. Better model parameters than those currently available for calculating the dose to the critical group can be obtained in this manner. Global dose estimates may require the development of techniques for estimating emissions occurring outside the growing season.

Contact Conductance between Cladding/Pressure Tube and Pressure Tube/Calandria Tube of Advanced Thermal Reactor (ATR)

In some postulated accidents with coincident loss of emergency coolant injection of the Advanced Thermal Reactor (ATR), the rate of heat transfer to the heavy water moderator that acts as heat sink for the decay heat depends on the contact conductance between cladding and pressure tube, and the same between pressure and calandria tubes. Experiments were performed to assess these contact conductances for clean plates and plates with simulated crud of Fe2O3 powder that is the main ingredient of the crud, and the applicable correlations were also studied. Test specimens were cut from actual pressure tube made of Zr-2.5%Nb and calandria tube made of Zircaloy-2 and flattened to sizes. The artificial waviness of various kinds of height and wave length of 10 mm was machined on the surface of the pressure tube specimen. The ranges of contact pressure, roughness, specimen temperature and gas pressure were from 0.5 to 7 MPa, 4.8 to 100 μm, 400 to 840 K and 0.001 Torr to atmospheric respectively. The experimental results without crud compare favorably with Tachibana's correlation. The experimental contact conductances with crud are found to be lowered. Since Tachibana's correlation was not applicable to the present situation, the empirical correlation was proposed. The correlation fitted well with the experimental values.

Microdosimetric Radiation Field Characterisation and Dosimetry in a Heavy Water Moderated Reactor Environment

Microdosimetric Radiation Field Characterisation and Dosimetry in a Heavy Water Moderated Reactor Environment

Microdosimetric Radiation Field Characterisation and Dosimetry in a Heavy Water Moderated Reactor Environment

Microdosimetric Radiation Field Characterisation and Dosimetry in a Heavy Water Moderated Reactor Environment

Contact Conductance between Cladding/Pressure Tube and Pressure Tube/Calandria Tube of Advanced Thermal Reactor (ATR).

In some postulated accidents with coincident loss of emergency coolant injection of the Advanced Thermal Reactor (ATR), the rate of heat transfer to the heavy water moderator that acts as heat sink for the decay heat depends on the contact conductance between cladding and pressure tube, and the same between pressure and calandria tubes. Experiments were performed to assess these contact conductances for clean plates and plates with simulated crud of Fe2O3 powder that is the main ingredient of the crud, and the applicable correlations were also studied. Test specimens were cut from actual pressure tube made of Zr-2.5%Nb and calandria tube made of Zircaloy-2 and flattened to sizes. The artificial waviness of various kinds of height and wave length of 10 mm was machined on the surface of the pressure tube specimen. The ranges of contact pressure, roughness, specimen temperature and gas pressure were from 0.5 to 7 MPa, 4.8 to 100 μm, 400 to 840 K and 0.001 Torr to atmospheric respectively. The experimental results without crud compare favorably with Tachibana's correlation. The experimental contact conductances with crud are found to be lowered. Since Tachibana's correlation was not applicable to the present situation, the empirical correlation was proposed. The correlation fitted well with the experimental values.

Core coolability of an ATR by heavy water moderator in situations beyond design basis accidents

In the case of a loss-of-coolant accident (LOCA) with coincident loss of emergency coolant injection (LOECI), core cooling is generally very severe. However, as the ATR plant has heavy water at about 60°C in the core, decay heat can be removed by the heavy water cooling system. Separate-effects tests relating to heavy water cooling were conducted with each setup. The important thermal hydraulics was radiation heat transfer, ballooning of a pressure tube, contact conductance between the pressure tube and a calandria tube and critical heat flux of the calandria tube. Constants and correlations obtained by the tests were incorporated into several codes to assess the core cooling. Long term core cooling capability with the heavy water cooling system was assessed. The core was cooled without melting under the postulated events due to inherent characteristics of the ATR.

Advanced technologies for CANDU reactors

The inherent features of a reactor based on multiple pressure tubes, rather than a single pressure vessel, provide CANDU with considerable flexibility for continuous design improvements. Of particular significance is the presence of a large volume of cool heavy-water moderator within the core. This can act as a heat sink for passive heat rejection during high-temperature accidents. We describe elements of our R&D program that are aimed to exploit this benefit. Similarly, safety, reliability and economy are the focus for our exploration of advanced technologies in the areas of advanced fuel and fuel channels, improved heavy-water production and computer-based systems that facilitate reactor design, construction and operation.

Target/Blanket Conceptual Designs for the Los Alamos Accelerator Transmutation of Nuclear Waste (ATW) Concept

Two target designs and one blanket design are discussed for the Los Alamos Accelerator Transmutation of Nuclear Waste concept. The blanket design presented includes heavy water moderator and piping systems used to contain fission products and actinide streams for transmutation to stable isotopes. A flowing liquid metal target offers the advantage of direct convection of the high heat generation and the possibility for a windowless design. Experiments will be required to test the target design and determine compatibility of the container material. Pumping and heat-transfer equipment will also require research and development. A solid target made of a combination of tungsten and lead offers the advantage of straightforward engineering design, but will require a window between the accelerator and the target.

Preliminary Risks Associated with Postulated Tritium Release from Production Reactor Operation

The Probabilistic Risk Assessment (PRA) of Savannah River Plant (SRP) reactor operation is evaluating the offsite risk due to tritium releases during postulated full or partial loss of heavy water moderator accidents. Preliminary determination of the frequency of average partial moderator loss (including incidents with leaks as small as 0.5 kg) yields an estimate of --1 per reactor-year. The full moderator loss frequency is conservatively chosen as 5x10/sup -3/ per reactor-year. Conditional consequences, determined with a version of the MACCS code modified to handle tritium, are found to be insignificant. The 95th percentile individual cancer risk is 2x10/sup -8/ per reactor-year within 16 km of the release point. The full moderator loss accident contributes about 80% of the evaluated risks.