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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Irradiation Dose of the Woody Tier of a Coniferous Forest Due to Accidental Emissions from NPP

An approach to the comparative assessment of the impact of accidental radioactive emissions on the biota and radioecological validation of nuclear fuel cycle objects is presented. It is shown that the dose loads on woody tier in a coniferous forest as a result of severe accidents in NPPs with PWR-890, BWR-1412, and EPR-1600 significantly exceed the limit recommended by the ICRP. The impact of the radiation factor can result in serious disturbances in forest ecosystems in the immediate vicinity of NPPs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI1992

Nuclear Safety [Vol. 33, No. 4, October-December 1992]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 499 Technical Note: Nuclear Regulatory Commission’s Examination Question Bank System, C. L Jensen and N. L. Skinner; 506 Report on the Specialists’ Meeting on Passive and Active Safety Features of Liquid-Metal Fast Breeder Reactors Organized by the International Atomic Energy Agency at Oarai Engineering Centre of Power Reactor and Nuclear Development Corporation, Japan, November 5-7,1991, S. R. Paranjpe; ACCIDENT ANALYSIS: 514 A Review of Hydrogen Production During Melt/Water Interaction in LWRs, D. F. Fletcher, B. D. Turland, and S. P. A. Lawrence; 535 A Review of the U 3 0 8 -Al Reaction as a Potential Heat Source in Research and Test Reactor Accidents, J. L. Snelgrove; CONTROL AND INSTRUMENTATION: 543 Multilevel Interfaces for Power Plant Control Rooms II: A Preliminary Design Space, K. J. Vicente; DESIGN FEATURES: 549 Twenty-Second DOE/NRC Nuclear Air Cleaning and Treatment Conference, R. R. Bellamy, D. W. Moeller, and M. W. First; 562 Book Review: Safety Assessment of Unit 5 (WWER-440/W-213) of the Greifswald Nuclear Power Station, M. H. Fontana; WASTE AND SPENT FUEL MANAGEMENT: 576 Activities Related to Waste and Spent Fuel Management, M. D. Muhlheim and E. G. Silver; OPERATING EXPERIENCES: 586 Analysis of Corrective Actions Applied to Nuclear Power Plant Operations, A. E. Cross; 593 Reactor Shutdown Experience, Compiled by J. W. Cletcher; 596 Operating U.S. Power Reactors, Compiled by M. D. Muhlheim and E. G. Silver; 609 Non-Commercial Power Operating Experience; 614 World Nuclear Performance, Compiled by M. D. Muhlheim; RECENT DEVELOPMENTS: 623 General Administrative Activities, Compiled by M. D. Muhlheim and E. G. Silver; 635 Reports, Standards, and Safety Guides, D. S. Queener; 639 Proposed Rule Changes as of June 30, 1992; ANNOUNCEMENTS: 513 Fourth International Conference on Nuclear Fuel Reprocessing and Waste Management RECOD '94; 585 International Topical Meeting on Advanced Reactors Safety; 622 International Workshop on Implementation of ALARA at Nuclear Power Plants; 648 Short Course on Nuclear Criticality Safety; 648 8th Topical Meeting on Problems in Reactor Physics MEPhi-VOLGA-93; 643 The Authors; 646 Reviewers of Nuclear Safety, Vol. 33

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI1989

Nuclear Safety [Vol. 30, No. 1, January-March 1989]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 1 International Conference on Thermal Reactor Safety: NUCSAFE 88 by A. P. Malinauskas and J. G. Pruett; ACCIDENT ANALYSIS: 18 Chernobyl Today: State of Research by A. Yu. Gagarinskii; CONTROL AND INSTRUMENTATION: 23 Problems Associated with Routine In-Plant Radioactive Effluent Monitoring Systems at U.S. Light-Water Reactors by Barry G. Wahlig, David M. Walker, Mahmoud R. Ghavi, and John M. Palms; DESIGN FEATURES: 34 Twentieth DOE/NRC Nuclear Air-Cleaning Conference by R. R. Bellamy, D. W. Moeller, and M. W. First, 46 Design Options for the Core Support Structure for Liquid-Metal-Cooled Reactor Plants by J. P. Burelbach, W. J. Kann, J. G. Saiveau, and R. W. Seidensticker; ENVIRONMENTAL EFFECTS: 53 The Radiological Consequences of the Chernobyl Accident by L. V. Konstantinov and A. J. González; WASTE AND SPENT FUEL MANAGEMENT: 70 Activities Related to Waste Management Compiled by E. G. Silver; OPERATING EXPERIENCES: 80 Fuel Performance Annual Report for 1986 by W. J. Bailey, 103 Safety Considerations Related to 12-Hour Shift Schedules by S. Gould, 105 Reactor Shutdown Experience Compiled by J. W. Cletcher, 108 Operating U.S. Power Reactors Compiled by E. G. Silver; RECENT DEVELOPMENTS: 130 General Administrative Activities Compiled by E. G. Silver, 140 Reports, Standards, and Safety Guides by D. S. Queener, 145 Status of Power-Reactor Licensing Activities Compiled by E. G. Silver, 153 Proposed Rule Changes as of Sept. 30, 1988; ANNOUNCEMENTS: 69 Northwestern University Short Course on Radiation Safety, 79 International Seminar on Fission Product Transport Processes in Reactor Accidents, 129 Harvard School of Public Health Offers Short Courses, 152 MIT Offers Summer Short Courses on Reactor Safety and Nodal Analysis, 152 Seminar on Use of PCs in Probabilistic Safety Analysis, 157 The Authors, 160 Indexes to Nuclear Safety, Volume 29.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Westinghouse Quiver - A simple and safe system for the complete handling and storing of failed BWR and PWR fuel rods and fragments - 17049

Background: Extracted leaking fuel rods from fuel assemblies have been stored in the spent fuel pools at the Nuclear Power Plants for many years. Sending those fuel rods to an intermediate storage has been very difficult due to the uncertainty of the technical requirements. The Westinghouse Quiver is a container that can be used from the first moment when the leaking fuel rod is extracted from the Fuel Assembly. The leaking fuel rod can be placed in the Westinghouse Quiver. When the Quiver is filled, it can be sealed and dried and then treated as a regular fuel assembly during intermediate storage and final repository. Purpose of System: The purpose of the system is to store leaking fuel rods, broken rods and fuel rod fragments leak tight in the fuel pool and after vacuum drying and permanent sealing at an intermediate storage facility. The Quiver is designed to be fully compatible with a regular fuel assembly. That means all handling of the Quiver in a Nuclear Power Plant, during transportation or storage is also identical to that of a regular fuel assembly. Fulfilling these criteria together with functional testing, the Quiver is considered acceptable for long time storage of leaking fuel rods, both in dry or wet condition. The Quiver has been qualified in Switzerland for use in the spent fuel pool, intermediate storage, and for a final repository. The system will be implemented as well in Sweden during 2016-2019. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Improvements in the reactor core isolation cooling (RCIC) pump model

As part of the efforts to understand the unexpected “self-regulating” mode of the RCIC (Reactor Core Isolation Cooling) systems that was observed during the Fukushima accidents and to extend BWR RCIC and PWR AFW (Auxiliary Feed Water) operational range and flexibility, mechanistic models for the Terry turbine are being developed and tested with the RELAP-7 code to simulate the RCIC system. In the previous work, we developed a set of analytical models for the normal working conditions, as well as the two-phase off-design conditions, of the RCIC system, based on Sandia’s original work. An under-expanded jet model was developed to replace the CFD-based reduced-order model in the Sandia work to obtain the velocity and thermodynamic conditions of the turbine stator inlet. The models include both an adiabatic expansion process inside the nozzle and a free expansion process outside of the nozzle to ambient pressure. For the two-phase off-design conditions, two well-established choking models were used - the Isentropic Homogenous Equilibrium Model (IHEM) and Moody’s model. The new twophase Terry turbine model used the choking models to calculate the mass flow rate, the critical pressure at the nozzle throat, and steam quality. The two-phase nozzle expansion model was also developed by considering the vapor phase with a similar model for the single-phase case with the assumption that the liquid phase would slip along the wall with a much slower speed and will not contribute to the impulse on the rotor.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

On the Solidification and Structure Formation during Casting of Large Inserts in Ferritic Nodular Cast Iron

The graphite nodule count and size distributions for boiling water reactor (BWR) and pressurized water reactor (PWR) inserts were investigated by taking samples at heights of 2160 and 1150 mm, respectively. In each cross section, two locations were taken into consideration for both the microstructural and solidification modeling. The numerical solidification modeling was performed in a two-dimensional model by considering the nucleation and growth in eutectic ductile cast iron. The microstructural results reveal that the nodule size and count distribution along the cross sections are different in each location for both inserts. Finer graphite nodules appear in the thinner sections and close to the mold walls. The coarser nodules are distributed mostly in the last solidified location. The simulation result indicates that the finer nodules are related to a higher cooling rate and a lower degree of microsegregation, whereas the coarser nodules are related to a lower cooling rate and a higher degree of microsegregation. The solidification time interval and the last solidifying locations in the BWR and PWR are also different.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Dancoff-based Wigner-Seitz approximation for the subgroup resonance self-shielding in the VERA neutronic simulator MPACT

The MPACT neutronics module of the Virtual Environment for Reactor Analysis (VERA) has used the subgroup method for resonance self-shielding calculation, for which two-dimensional (2D) fixed-source transport calculations are performed using the method of characteristics for resonance energy groups. When considering thermal feedbacks, the subgroup calculation must be performed at each outer iteration. Therefore, the computing time for cross section processing is a significant burden for computational efficiency. The Dancoff-based Wigner-Seitz approximation (DWA) capability has been implemented into MPACT in conjunction with the subgroup method, which has been used in SCALE/XSProc since SCALE version 6.0 and has recently been called an equivalent Dancoff-factor cell (EDC) method. The issue of relatively large reactivity bias in DWA for the gadolinia rods was resolved by introducing multiple Dancoff factors. Benchmark results for the VERA pressurized and boiling water reactor benchmark suites show that the DWA capability would significantly enhance computational efficiency with comparable accuracy. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Multiscale thermal hydraulic coupling methods for boiling water reactor simulation

During the last two years, the Virtual Environment for Reactor Applications (VERA) has been extended to simulate boiling water reactors (BWRs). The thermal hydraulic effects present in BWRs are far more complex than those in pressurized water reactors (PWRs). Therefore, the runtime is significantly increase and convergence behavior worse compared to PWRs. Most of the additional runtime is spent during the first few coupled iterations, when the power shape is still rapidly evolving, dramatically affecting the thermal hydraulics (TH). To alleviate the increased computational expense, a multiscale TH coupling approach was developed in VERA in which a highly efficient, simplified TH model is solved for several coupled iterations until the power shape is partially converged. The simplified solution is used to inform the assembly-wise flow distribution in the high-fidelity TH solver CTF, reducing the amount of work required to properly balance the pressure drop in each channel and reducing runtime for the couple calculation. Because the final TH calculations are performed with CTF, there is ultimately no impact on the accuracy of the converged solution. This paper presents the details of this multiscale TH coupling approach, along with results for a 4x4 array of GE-14 fuel bundles and whole-core coupled simulations of the Hatch core. These two cases show that the multiscale approach dramatically reduces the runtime of coupled BWR simulations. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Multi-phase model development to assess RCIC system capabilities under severe accident conditions

The Reactor Core Isolation Cooling (RCIC) System is a safety-related system that provides makeup water for core cooling of some Boiling Water Reactors (BWRs) with a Mark I containment. The RCIC System consists of a steam-driven Terry turbine that powers a centrifugal, multi-stage pump for providing water to the reactor pressure vessel. The Fukushima Dai-ichi accidents demonstrated that the RCIC System can play an important role under accident conditions in removing core decay heat. The unexpectedly sustained, good performance of the RCIC System in the Fukushima reactor demonstrates, firstly, that its capabilities are not well understood, and secondly, that the system has high potential for extended core cooling in accident scenarios. Better understanding and analysis tools would allow for more options to cope with a severe accident situation and to reduce the consequences. The objectives of this project were to develop physics-based models of the RCIC System, incorporate them into a multi-phase code and validate the models. This Final Technical Report details the progress throughout the project duration and the accomplishments.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Quantification of functional impact classification on the current U.S. nuclear fleet

Currently, nuclear power generates only a small fraction of the world's electricity, though it does have the potential to meet the needs of the world as long as its production can compete with the alternative energy base load producers. In U.S the commercial nuclear fleet comprises of Light Water Reactors (LWRs), consisting of Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). Many plants experience multiple events each year, which lead to reactor trips or scrams. This subsequently leads to plants being off the grid, which is not an economically favorable situation in the current energy markets. This study helps identify functional impact (FI) event categories for both PWR's and BWR's. Understanding that not all events and their causes can be anticipated, current industry best practices implement lessons learned to provide challenges and issues identification. In turn these are studied to achieve improved operational practices, and design modifications. This leads to overall enhanced resiliency. FI groups are the categories of initiating events (IE). IEs are unplanned events that occur while a nuclear plant is in critical operation, and requires that the plant be shut down to achieve a stable state. When a risk cannot be eliminated, IEs are identified and studied to reduce the risk and contribute to the identification of mitigations, safer designs and operational practices. Breaking IE events into groups' results in the classification of FI groups. A FI group is a risk-significant event category that could impact the ability of a NPP to remove decay heat from the reactor. Inability to remove decay heat from the reactor has the potential to lead to core damage. For an event to be considered a FI, it must be associated with a manual or automatic reactor trip, irrespective of the order of events. In addition, the plant must be at or above the point of adding heat and the event must happen before or shortly after a reactor trip. One or more FIs may be identified in any individual reactor trip sequence. The relationship between FI sequence, and/or frequency per a plant event is outside the scope of this paper. For example, the loss of offsite power (LOOP) can occur from eternal events (i.e. weather, human, wildlife etc.). This results in an event that requires a transfer of power to an emergency source, such as a generator, which produces a reactor trip and then a closure of main steam isolation valves (MSIVS). The FI applicable to this reactor trip sequence would be LOOP and closure of at least one MSIV in each main steam line. The results presented herein are focused on the analysis and evaluation of FI event categories. Only FI categories that can be found in IE studies, Initialing Events Spreadsheet and are recorded in a historical review of Licensee Event Reports (LERs) from 1988 through 2013. The results of this analysis can support plant decisions, design, procedures, and identify weakness.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Native Natural circulation in VERA

This work seeks to introduce a native natural circulation solver to VERA's thermohydraulics capabilities. This work begins with initial investigations into the viability of such a solver by determining the fundamental equations as well as necessary inputs. The method was then implemented in the form of a Python script coupled to CTF which was compared to TRACE. Once viability of the natural circulation solver was established, it was then implemented natively in MPACT for use with previously created simplified thermohydraulics solvers. This new implementation of the native natural circulation solver in MPACT is then tested for single physics problems for both a BWR and PWR single channel model and compared to TRACE where it shows an error of about 30% for BWRs and about 15% for PWRs. The solver is additionally tested for NuScale based models where it demonstrates an error to the reported flow rate of about 10%, resulting in an eigenvalue error of only 138 pcm when comparing fixed flow rate multiphysics runs. Finally, the solver is tested in multiphysics systems for a small PWR assembly, a small BWR assembly, and a full core approximate model of the NuScale design. In these multiphysics runs, it demonstrates no impact on multiphysics convergence stability, as well as negligible impact on runtimes. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

P{sub N} source expansion nodal method in MPACT for boiling water reactors

This paper describes the one-node P{sub N}-Source Expansion Nodal Method (SENM) axial solver for the 2D/1D method recently implemented in MPACT to support Boiling Water Reactors (BWR) analysis. Since the BWR has a more complicated design and strong burnable absorber, the existing PN-Nodal Expansion Method (NEM) axial solver in MPACT may not be sufficient to accurately represent the intranodal flux and source profiles for BWRs. The one-node P{sub N}-SENM has been implemented in this work to reduce the axial spatial discretization error for BWRs. From numerical results, we verify that the P{sub N}-SENM can improve the accuracy of the pin power prediction, and confirm that P{sub N}-SENM can use more than a 1.5 times larger axial mesh size than P{sub N}-NEM to have similar accuracy for the BWR GE14 3D assembly problem. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

High-burnup isotopics bias and uncertainty estimations for LANCR02

Calculated nuclide concentrations are compared against destructive radiochemical assays of modern, high-burnup BWR fuel to validate the LANCR02 lattice physics code. Good agreement is observed between different lattice physics methodologies - including the LANCR02 implementation of Method of Characteristics neutron and gamma transport solver and the TGBLA06 methodology - and against measurements. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

VERA BWR progression problems

During the first phase of the Consortium for Advanced Simulation of Light Water Reactors (CASL) program, the Virtual Environment for Reactor Applications (VERA) was developed with a focus on capabilities for high-fidelity, multiphysics simulation of pressurized water reactors (PWRs). During this development effort, a set of progression problems was created ranging from smaller pin cell calculations to larger 3D full-core calculations. These progression problems helped to guide the development of the software and served as benchmarks against which to test VERA. Since 2019, efforts have been made to extend the capabilities of VERA to model boiling water reactors (BWRs). Because BWR simulations come with many unique challenges, a set of BWR progression problems was developed to aid in this new effort. The BWR progression problems range from 2D lattice calculations to 3D mini-core problems, and reference neutronic solutions were computed using continuous-energy Monte Carlo codes. MPACT, one of the neutronics code in VERA, was benchmarked using the BWR progression problems. The code is capable of computing solutions to all problems. The eigenvalues computed by MPACT agree well with the Monte Carlo reference solutions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

External Cooling of the BWR Mark I and II Drywell Head as a Potential Accident Mitigation Measure – Scoping Assessment

This report documents a scoping assessment of a potential accident mitigation action applicable to the US fleet of boiling water reactors with Mark I and II containments. The mitigation action is to externally flood the primary containment vessel drywell head using portable pumps or other means. A scoping assessment of the potential benefits of this mitigation action was conducted focusing on the ability to (1) passively remove heat from containment, (2) prevent or delay leakage through the drywell head seal (due to high temperatures and/or pressure), and (3) scrub radionuclide releases if the drywell head seal leaks.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

SiC Composite for Fuel Structure Applications

Extensive evaluation was performed to determine the suitability of using SiC composite as a boiling water reactor (BWR) fuel channel material. A thin walled SiC composite box, 10 cm in dimension by approximately 1.5 mm wall thickness was fabricated using chemical vapor deposition (CVD) for testing. Mechanical test results and performance evaluations indicate the material could meet BWR channel mechanical design requirement. However, large mass loss of up to 21% was measured in in-pile corrosion test under BWR-like conditions in under 3 months of irradiation. A fresh sister sample irradiated in a follow-up cycle under PWR conditions showed no measureable weight loss and thus supports the hypothesis that the oxidizing condition of the BWR-like coolant chemistry was responsible for the high corrosion rate. A thermodynamic evaluation showed SiC is not stable and the material may oxidize to form SiO 2 and CO 2 . Silica has demonstrated stability in high temperature steam environment and form a protective oxide layer under severe accident conditions. However, it does not form a protective layer in water under normal BWR operational conditions due to its high solubility. Corrosion product stabilization by modifying the SiC CVD surface is an approach evaluated in this study to mitigate the high corrosion rate. Titanium and zirconium have been selected as stabilizing elements since both TiSiO 4 and ZrSiO 4 are insoluble in water. Corrosion test results in oxygenated water autoclave indicate TiSiO4 does not form a protective layer. However, zirconium doped test samples appear to form a stable continuous layer of ZrSiO 4 during the corrosion process. Additional process development is needed to produce a good ZrSiC coating to verify functionality of the mitigation concept.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Impact of assembly-specific conditions on BWR burnup credit

Because boiling water reactors use control blades during operation, there can be significant changes to the local axial power shape, coolant density profile, and other parameters when the blades are inserted. Previous work examined the reactor operating parameters that impact cask reactivity. These previous studies identified limiting conditions for fuel cask reactivity or the direction of trends with cask reactivity (e.g., lower coolant densities result in higher cask reactivity). However, using limiting conditions for all parameters simultaneously is unrealistic. When the control blades are inserted deeply into the reactor - a limiting condition - the power is reduced and the void fraction decreases - a less limiting condition. This paper identifies the impacts of using assembly-specific conditions for control blade history, coolant density profile, burnup profile, and fuel temperature profile. Results indicate that cask reactivity is reduced when using assembly-specific operating conditions versus combining limiting conditions for the individual parameters of interest. The magnitude of reactivity reduction for using assembly- specific conditions varies based on each assembly and its operating conditions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Sensitivity and uncertainty analysis for a combined depletion and criticality calculation chain of a BWR FA

This paper examines the prerequisites and boundary conditions for extending burnup credit from pressurized water reactors to boiling water reactors with validated depletion codes by performing uncertainty and sensitivity analyses on a modern boiling water reactor fuel assembly design. We developed a computational chain consisting of an axial boiling water reactor axial irradiation model, followed by a series of depletion calculations and a three-dimensional criticality calculation. The irradiation model puts into relation the profiles of power, burnup and moderator density of the fuel assembly at 24 axial locations. In a simplified fashion it also includes the insertion of control blades. The depletion calculations of the fuel assembly were performed in axial slices with the 2D-depletion code HELIOS 1.12. The criticality calculations were done in 3D-geometry with KENO-Va from the SCALE 6.2.1 package, in which several fuel assemblies are placed into a generic transport cask. For the uncertainty and sensitivity analyses a Monte Carlo sampling based approach is used applying the GRS tool SUnCISTT. We performed several generic investigations testing the capabilities and the limitations of the whole computational chain. The impact of variations of the irradiation model on the resulting reactivity of the fuel assemblies inside a transport cask or storage pool was investigated using an uncertainty and sensitivity analysis on the whole computational chain. Geometrical uncertainties were also included in the evaluation and a parameter study on the use of control blades during irradiation was performed. The goal was to develop the procedures and tools to be able to obtain quantitative uncertainty data which can be used in the future to derive conservative irradiation profiles without unnecessary excess reactivity margins. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗