Consistent Data Assimilation of Isotopes: 242Pu and 105Pd

In this annual report we illustrate the methodology of the consistent data assimilation that allows to use the information coming from integral experiments for improving the basic nuclear parameters used in cross section evaluation. A series of integral experiments are analyzed using the EMPIRE evaluated files for 242Pu and 105Pd. In particular irradiation experiments (PROFIL-1 and -2, TRAPU-1, -2 and -3) provide information about capture cross sections, and a critical configuration, COSMO, where fission spectral indexes were measured, provides information about fission cross section. The observed discrepancies between calculated and experimental results are used in conjunction with the computed sensitivity coefficients and covariance matrix for nuclear parameters in a consistent data assimilation. The results obtained by the consistent data assimilation indicate that not so large modifications on some key identified nuclear parameters allow to obtain reasonable C/E. However, for some parameters such variations are outside the range of 1 s of their initial standard deviation. This can indicate a possible conflict between differential measurements (used to calculate the initial standard deviations) and the integral measurements used in the statistical data adjustment. Moreover, an inconsistency between the C/E of two sets of irradiation experiments (PROFIL and TRAPU) is observed for 242Pu. This is the end of this project funded by the Nuclear Physics Program of the DOE Office of Science. We can indicate that a proof of principle has been demonstrated for a few isotopes for this innovative methodology. However, we are still far from having explored all the possibilities and made this methodology to be considered proved and robust. In particular many issues are worth further investigation: • Non-linear effects • Flexibility of nuclear parameters in describing cross sections • Multi-isotope consistent assimilation • Consistency between differential and integral experiment

On Perturbation Components Correspondence between Diffusion and Transport

We have established a correspondence between perturbation components in diffusion and transport theory. In particular we have established the correspondence between the leakage perturbation component of the diffusion theory to that of the group self scattering in transport theory. This has been confirmed by practical applications on sodium void reactivity calculations of fast reactors. Why this is important for current investigations? Recently, there has been a renewed interest in designing fast reactors where the sodium void reactivity coefficient is minimized. In particular the ASTRID8,9 reactor concept has been optimized with this goal in mind. The correspondence on the leakage term that has been established here has a twofold implication for the design of this kind of reactors. First, this type of reactor has a radial reflector; therefore, as shown before, the sodium void reactivity coefficient calculation requires the use of transport theory. The minimization of the sodium reactivity coefficient is normally done by increasing the leakage component that has a negative sign. The correspondence established in this paper allows to directly look at this component in transport theory. The second implication is related to the uncertainty evaluation on sodium void reactivity. As it has shown before, the total sodium void reactivity effect is the result of a large compensation (opposite sign) between the scattering (called often spectral) component and the leakage one. Consequently, one has to evaluate separately the uncertainty on each separate component and then combine them statistically. If one wants to compute the cross section sensitivity coefficients of the two different components, the formulation established in this paper allows to achieve this goal by playing on the contribution to the sodium void reactivity coming from the group self scattering of the sodium cross section

Anisotropic scattering in the variational nodal simplified spherical harmonics formulation

Under the assumption of isotropic scattering, the simplified spherical harmonics method (SP{sub N}) was recently formulated in variational nodal form and implemented successfully as an option of the VARIANT code. The authors here remove the isotopic scattering restriction. The variational nodal form of the SPN approximation is formulated and implemented with both within-group and group-to-group anisotropic scattering. Results are presented for a model problem previously utilized with the standard P{sub N} variational nodal method

Comparison of simplified and standard spherical harmonics in the variational nodal method

Recently, the variational nodal method has been extended through the use of the Rumyantsev interface conditions to solve the spherical harmonics (P{sub N}) equations of arbitrary odd order. In this paper, the authors generalize earlier x-y geometry work to fit the corresponding simplified spherical harmonics (SP{sub N}) equations into the variational nodal framework. Both P{sub N} and SP{sub N} approximations are implemented in the multigroup VARIANT code at Argonne National Laboratory in two and three dimensional Cartesian and hexagonal geometries. The availability of angular approximations through P{sub 5} and SP{sub 5}, and of flat, linear and quadratic spatial interface approximations allows investigation of both spatial truncation and angular approximation errors. Moreover, the SP{sub 3} approximation offers a cost-effective method for reducing transport errors

Sensitivity and representativity analysis of past experiments with respect to ABTR system.

A comprehensive validation analysis has been performed that incorporates representativity of multiple parameters, experiments, reference designs, and adjustment of the nuclear data. The work involves a new representativity study among selected reactor designs and several experiments. Application, using existing experiments, to reference design like the ABTR and the SFR has demonstrated that it is possible to achieve a significant reduction of uncertainty on the main integral parameters of interest for their neutronic design. This is possible when the set of available experiments are relevant (i.e. representative of the reference designs), of good quality (i.e. of reduced uncertainty on experimental results), and consistent (i.e. not providing conflictive information)

Simultaneous Nuclear Data Target Accuracy Study for Innovative Fast Reactors

The present paper summarizes the major outcomes of a study conducted within a Nuclear Energy Agency Working Party on Evaluation Cooperation (NEA WPEC) initiative aiming to investigate data needs for future innovative nuclear systems, to quantify them and to propose a strategy to meet the

Pathologic Complete Response in Urothelial Carcinoma Patients Receiving Neoadjuvant Immune Checkpoint Inhibitors: A Meta-Analysis

Background. Immune checkpoint inhibitors (ICIs) have been evaluated as neoadjuvant treatment in urothelial carcinoma (UC) patients, with these agents reporting encouraging pathologic complete response (pCR) rates. Herein, we performed a systematic review and meta-analysis aimed at evaluating the incidence of pCR in UC patients treated with neoadjuvant ICI. Moreover, we investigated the impact of PD-L1 expression in this patient population, exploring the possible role of PD-L1 status as predictive biomarker. Materials and Methods. We retrieved all the relevant trials through PubMed/Medline, Cochrane Library and EMBASE; moreover, proceedings of the main international oncological meetings were also searched for relevant abstracts. Eligible trials assessed pre-operative ICI in UC patients. Results. Our meta-analysis has highlighted a pooled pCR rate of 36.6% in the overall population; interestingly, pCR was higher in PD-L1 positive compared with PD-L1 negative UCs (49.5% versus 35.1%, respectively). Conclusions. Positive signals emanating from neoadjuvant immunotherapy should encourage the scientific community to persist in the long road toward finding more effective treatments for UC patients

Validation of Simulation Codes for Future Systems: Motivations, Approach and the Role of Nuclear Data

The validation of advanced simulation tools will still play a very significant role in several areas of reactor system analysis. This is the case of reactor physics and neutronics, where nuclear data uncertainties still play a crucial role for many core and fuel cycle parameters. The present paper gives a summary of validation motivations, objectives and approach. A validation effort is in particular necessary in the frame of advanced (e.g. Generation-IV or GNEP) reactors and associated fuel cycles assessment and design

Target Accuracy Assessment for an ADS Design

Nuclear data uncertainties and their impact on a very wide range of reactor systems, including their associated fuel cycles, have to be assessed in order to consolidate preliminary design studies for new innovative systems. One specific class of systems is the so-called “dedicated waste transmuters”, that are fast neutron systems (critical or sub-critical, i.e. ADS), loaded with a Minor Actinide (MA) dominated fuel and potentially uranium-free. The availability of very general tools for sensitivity and uncertainty analysis together with new variance-covariance matrix data, produced in a joint effort under the auspices of the OECD-NEA by the world leading nuclear data evaluation groups, makes that endeavor particularly significant. In this report major results of interest for dedicated ADS are discussed and the most important fields and data types are pointed out, where priority improvements are required

Methods in Use for Sensitivity Analysis, Uncertainty Evaluation, and Target Accuracy Assessment

Sensitivity coefficients can be used for different objectives like uncertainty estimates, design optimization, determination of target accuracy requirements, adjustment of input parameters, and evaluations of the representativity of an experiment with respect to a reference design configuration. In this paper the theory, based on the adjoint approach, that is implemented in the ERANOS fast reactor code system is presented along with some unique tools and features related to specific types of problems as is the case for nuclide transmutation, reactivity loss during the cycle, decay heat, neutron source associated to fuel fabrication, and experiment representativity