MCNP: Multigroup/adjoint capabilities

Wagner, John C.; Redmond, E.L.; Palmtag, Scott; Hendricks, John S.

doi:10.2172/10145615

Cited by 30 publications

(18 citation statements)

References 2 publications

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“…It should be noted that MCNP adjoint results have to be normalized to the real treatment source. This normalizing is described by Wagner et al 16 and Difilippo. 6 A thorough description of adjoint MC for neutrons can be found in Lewis and Miller 17 and Hoogenboom.…”

Section: B Relation Between Forward Flux and Adjoint Functionmentioning

confidence: 82%

Application of adjoint Monte Carlo to accelerate simulations of mono‐directional beams in treatment planning for Boron Neutron Capture Therapy

et al. 2007

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This paper deals with the application of the adjoint transport theory in order to optimize Monte Carlo based radiotherapy treatment planning. The technique is applied to Boron Neutron Capture Therapy where most often mixed beams of neutrons and gammas are involved. In normal forward Monte Carlo simulations the particles start at a source and lose energy as they travel towards the region of interest, i.e., the designated point of detection. Conversely, with adjoint Monte Carlo simulations, the so-called adjoint particles start at the region of interest and gain energy as they travel towards the source where they are detected. In this respect, the particles travel backwards and the real source and real detector become the adjoint detector and adjoint source, respectively. At the adjoint detector, an adjoint function is obtained with which numerically the same result, e.g., dose or flux in the tumor, can be derived as with forward Monte Carlo. In many cases, the adjoint method is more efficient and by that is much quicker when, for example, the response in the tumor or organ at risk for many locations and orientations of the treatment beam around the patient is required. However, a problem occurs when the treatment beam is mono-directional as the probability of detecting adjoint Monte Carlo particles traversing the beam exit ͑detector plane in adjoint mode͒ in the negative direction of the incident beam is zero. This problem is addressed here and solved first with the use of next event estimators and second with the application of a Legendre expansion technique of the angular adjoint function. In the first approach, adjoint particles are tracked deterministically through a tube to a ͑adjoint͒ point detector far away from the geometric model. The adjoint particles will traverse the disk shaped entrance of this tube ͑the beam exit in the actual geometry͒ perpendicularly. This method is slow whenever many events are involved that are not contributing to the point detector, e.g., neutrons in a scattering medium. In the second approach, adjoint particles that traverse an adjoint shaped detector plane are used to estimate the Legendre coefficients for expansion of the angular adjoint function. This provides an estimate of the adjoint function for the direction normal to the detector plane. In a realistic head model, as described in this paper, which is surrounded by 1020 mono-directional neutron/gamma beams and from which the best ones are to be selected, the example calculates the neutron and gamma fluxes in ten tumors and ten organs at risk. For small diameter beams ͑5 cm͒, and with comparable relative errors, forward Monte Carlo is seen to be 1.5 times faster than the adjoint Monte Carlo techniques. For larger diameter neutron beams ͑10 and 15 cm͒, the Legendre technique is found to be 6 and 20 times faster, respectively. In the case of gammas alone, for the 10 and 15 cm diam beams, both adjoint Monte Carlo Legendre and point detector techniques are respectively 2 and 3 times faster than forward Monte Carlo.

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Section: B Relation Between Forward Flux and Adjoint Functionmentioning

confidence: 82%

Application of adjoint Monte Carlo to accelerate simulations of mono‐directional beams in treatment planning for Boron Neutron Capture Therapy

et al. 2007

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“…Without knowledge, the weight window would cause splitting and Russian roulette of particles unsuitably. The weight window generator 7) or the adjoint importance function generator 8) makes their parameter automatically even though the energy spectrum of the particle is unknown.…”

Section: Trunnion Region Geometry (1) Experimental Setup and Resultsmentioning

confidence: 99%

“…In the present study, the adjoint importance function generator, 8) which is used for the generation of adjoint importance functions in MCNP, is employed for the variance reduction technique. The cross section library used for the adjoint calculation is MENDF5 8) group constant generated from the code CRSRD.…”

Section: Trunnion Region Geometry (1) Experimental Setup and Resultsmentioning

confidence: 99%

“…The cross section library used for the adjoint calculation is MENDF5 8) group constant generated from the code CRSRD. 8) To do effective Monte Carlo analysis for the experiments, the superimposed mesh tally 3) was used for spatial dose-equivalent rate distribution and the point detector was used for the dose-equivalent rate on measurement point with rem counter. All of the fractional standard deviations in these calculations are within 0.05.…”

Section: Trunnion Region Geometry (1) Experimental Setup and Resultsmentioning

confidence: 99%

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Application of Dose Evaluation of the MCNP Code for the Spent Fuel Transport Cask

Asami

Sawada

Konnai

et al. 2011

Progress in Nuclear Science and Technology

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The spent fuel transport cask has structure based mostly on the multi-layer for the radiation shielding. However, the structure for neutron shielding is very weak around the trunnion. It is important to evaluate realistic dose-equivalent rate in shielding design of the spent fuel transport cask, therefore the three dimensional continuous-energy Monte Carlo radiation transport code that exactly treating the complicated geometry was applied. In the fixed source problem such as a neutron deep penetration calculation with the Monte Carlo method, the application of the variance reduction method is the most important for a high figure of merit and the most reliable calculation. The concerned items are setting method for the variance reduction. The validation of dose-equivalent rate evaluation for the spent fuel transport cask by Monte Carlo code were performed by two kinds of neutron shielding benchmark experiments, deep penetration experiment for the side surface of the cask and streaming experiment around the trunnion. Dose-equivalent rate distributions at each benchmark were measured and compared with the calculated results. The comparison showed a good consistency between calculation and experiment results.

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“…13) and CRSRD (Ref. 14). A forward neutron flux spectrum is required for multi-group weighting of quantities.…”

Section: Methodsmentioning

confidence: 99%

Adjoint-Based Uncertainty Analysis for Essential Reactions in a Laser Inertial Fusion Engine

Seifried

Fratoni

Kramer

et al. 2011

Fusion Science and Technology

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MCNP: Multigroup/adjoint capabilities

Cited by 30 publications

References 2 publications

Application of adjoint Monte Carlo to accelerate simulations of mono‐directional beams in treatment planning for Boron Neutron Capture Therapy

Application of adjoint Monte Carlo to accelerate simulations of mono‐directional beams in treatment planning for Boron Neutron Capture Therapy

Application of Dose Evaluation of the MCNP Code for the Spent Fuel Transport Cask

Adjoint-Based Uncertainty Analysis for Essential Reactions in a Laser Inertial Fusion Engine

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