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Analysis of MSIV-ATWS events with the BNL plant analyzer

Description: There are automatic safety features and operator-initiated emergency procedures which influence the sequence of events until the time when the standby liquid control system (SLCS), or other attempts to get control rods inserted, can effect shutdown of the core. One emergency procedure for a BWR/4 would require the operator to reduce the flow of high pressure coolant injection (HPCI) into the reactor. The core inlet flow rate at this time would be due to natural circulation and the reduced flow would lower the water level in the downcomer thereby reducing the natural circulation flow rate. This effect, and the reduction in core inlet subcooling due to mixing of the emergency feedwater with steam in the downcomer when the level was lowered, cause a sufficient increase in core void fraction so that the power would be reduced. A reduction in pressure might also be called for during this event in order to comply with the PSP heat capacity temperature limit (or possibly to prevent cycling of relief valves). In the past few years there have been several studies of this problem with the emphasis on calculating the power level in the core. In the present study we consider the power level as well as the resulting PSP temperature and take into account different assumptions regarding plant parameters and operator actions.
Date: June 1, 1986
Creator: Diamond, D.J.
Partner: UNT Libraries Government Documents Department

Effect of reactor conditions on MSIV (main steam isolation valves)-ATWS power level

Description: In a boiling water reactor (BWR) when there is closure of the main steam isolation valves (MSIVs), the energy generated in the core will be transferred to the pressure suppression pool (PSP) via steam flows out of the relief valves. The pool has limited capacity as a heat sink and hence, if there is no reactor trip (an ATWS event), there is the possibility that the pool temperature may rise beyond acceptable limits. The present study was undertaken to determine how the initial reactor conditions affect the power during an MSIV-ATWS event. The time of interest is during the 20-30 minute period when it is assumed that the reactor is in a quasi-equilibrium condition with the water level and pressure fixed, natural circulation conditions and no control rod movement or significant boron in the core. The initial conditions of interest are the time during the cycle and the operating state. 4 refs., 2 tabs.
Date: January 1, 1987
Creator: Diamond, D.J.
Partner: UNT Libraries Government Documents Department

Light-water-reactor coupled neutronic and thermal-hydraulic codes

Description: An overview is presented of computer codes that model light water reactor cores with coupled neutronics and thermal-hydraulics. This includes codes for transient analysis and codes for steady state analysis which include fuel depletion and fission product buildup. Applications in nuclear design, reactor operations and safety analysis are given and the major codes in use in the USA are identified. The neutronic and thermal-hydraulic methodologies and other code features are outlined for three steady state codes (PDQ7, NODE-P/B and SIMULATE) and four dynamic codes (BNL-TWIGL, MEKIN, RAMONA-3B, RETRAN-02). Speculation as to future trends with such codes is also presented.
Date: January 1, 1982
Creator: Diamond, D.J.
Partner: UNT Libraries Government Documents Department

Reactivity parameters for safety analysis

Description: The reactor core model in the most commonly used computer programs for safety analysis is a point kinetics model. The core average fission rate is calculated knowing the reactivity, neutron generation time and delayed-neutron parameters. The reactivity is a time dependent function taking account of the effect of changes in water density and temperature, fuel temperature, control rod position and soluble boron concentration. In this presentation some of the alternative ways of representing this reactivity function are reviewed.
Date: January 1, 1985
Creator: Diamond, D.J.
Partner: UNT Libraries Government Documents Department

Uncertainty in BWR power during ATWS events

Description: A study was undertaken to improve our understanding of BWR conditions following the closure of main steam isolation valves and the failure of reactor trip. Of particular interest was the power during the period when the core had reached a quasi-equilibrium condition with a natural circulation flow rate determined by the water level in the downcomer. Insights into the uncertainity in the calculation of this power with sophisticated computer codes were quantified using a simple model which relates power to the principal thermal-hydraulic variables and reactivity coefficients; the latter representing the link between the thermal-hydraulics and the neutronics. Assumptions regarding the uncertainty in these variables and coefficients were then used to determine the uncertainty in power.
Date: January 1, 1986
Creator: Diamond, D.J.
Partner: UNT Libraries Government Documents Department

Xenon changes under power-burst conditions. [BWR]

Description: Under ordinary operating conditions the xenon concentration in a reactor core can change significantly in times on the order of hours. Core transients of safety significance are much more rapid and hence calculations are done with xenon concentration held constant. However, in certain transients (such as reactivity initiated accidents) there is a very large power surge and the question arises as to whether under these circumstances the xenon concentration could change. This would be particularly important if the xenon were reduced thereby tending to make the accident autocatalytic. The objective of the present study is to quantify this effect to see if it could be important.
Date: January 1, 1983
Creator: Diamond, D.J.
Partner: UNT Libraries Government Documents Department

Use of one delayed-neutron precursor group in transient analysis. [PWR; BWR]

Description: In most reactor dynamics calculations six groups of delayed-neutron precursors are usually accounted for. However, under certain circumstances it may be advantageous to simplify the calculation and utilize a single delayed-neutron group. The motivation for going to one precursor group is economy. For LWR transient codes that use point kinetics the equations are solved very rapidly and six precursor groups should always be used. However, codes with spatially dependent neutron kinetics are very long running and the use of one precursor group may save computer costs and not impair the accuracy of the results significantly. Furthermore, in some codes, the elimation of five presursor groups makes additional memory available which may be used to give a net increase in the accuracy of the calculations, e.g., by allowing for an increase in mesh density. In order to use one delayed neutron precursor group it is necessary to derive a single decay constant, /sub 6/lambda-, which, along with the total (or one group) delayed neutron fraction ..beta.. = ..sigma../sub i = 1/..beta../sub i/, will adequately describe the transeint precursor behavior. The present summary explains how a recommendation for lambda- was derived.
Date: January 1, 1983
Creator: Diamond, D.J.
Partner: UNT Libraries Government Documents Department

Requirements for Reactor Physics Design

Description: It has been recognized that there is a need for requirements and guidance for design and operation of nuclear power plants. This is becoming more important as more reactors are being proposed to be built. In parallel with activities in individual countries are norms established by international organizations. This paper discusses requirements/guidance for neutronic design and operation as promulgated by the U.S. Nuclear Regulatory Commission (NRC). As an example, details are given for one reactor physics parameter, namely, the moderator temperature reactivity coefficient. The requirements/guidance from the NRC are discussed in the context of those generated for the International Atomic Energy Agency. The requirements/guidance are not identical from the two sources although they are compatible.
Date: April 11, 2008
Creator: Diamond,D.J.
Partner: UNT Libraries Government Documents Department

Calculation of MCPR (minimum critical power ratio) for BWR transients using the BNL plant analyzer

Description: The critical power ratio (CPR) is used for determining the thermal limits of boiling water reactors. In this study, critical power ratios for a series of transients run on the Brookhaven Plant Analyzer (BPA) (1) have been calculated. The transients include nominal base case simulations, simulations with variations in relief valve setpoints and the number of failed feedwater heaters, simulations at the 100% power, 75% flow point on the extended load line of the MEOD, and a simulation with partial feedwater heating. The plant represented with the BPA is a BWR/4 rated at 3293 MW with a 6.38 m (251'') vessel. Data were obtained by the Plant Analyzer Development Group at BNL from a variety of sources describing the Browns Ferry Plant.
Date: June 1, 1987
Creator: Horak, W.C. & Diamond, D.J.
Partner: UNT Libraries Government Documents Department

Model for boron transport and reactivity in a BWR

Description: RAMONA-3B, a code for calculating BWR transients, models the components of the nuclear steam supply system. This document describes the boron transport and reactivity model which has been added to the code in order to represent the standby liquid control system.
Date: January 1, 1981
Creator: Garber, D.I. & Diamond, D.J.
Partner: UNT Libraries Government Documents Department

Effect of thermal-hydraulic feedback on the BWR rod drop accident

Description: An important design-basis accident for boiling water reactors (BWR's) is the rod drop accident (RDA). This accident is defined to be a rapid reactor transient caused by an accidental drop (out of the core) of the highest-worth control rod at various conditions ranging from cold start-up to about 10% of rated power. For most BWR designs the highest worth rod is normally situated at the center of the core. Despite the fact that the chance of a RDA in extremely unlikely, the consequence of the RDA is of concern because of the potential for damage to fuel rods. Neglecting moderator feedback during the RDA is a poor assumption because energy is deposited in the fuel over a 3 to 4 second time period and hence there is time for heat to be conducted to the coolant. This may tend to ameliorate the accident considerably. Evaluation of the thermal-hydraulic feedback effect on the RDS in a BWR has been scarce in the literature. The object of this paper is to demonstrate the beneficial effect of thermal-hydraulic feedback in the RDA.
Date: January 1, 1979
Creator: Cheng, H.S. & Diamond, D.J.
Partner: UNT Libraries Government Documents Department

BEAGL-01, a computer code for calculating rapid LWR core transients

Description: BEAGL-01 (Brookhaven's and EPRI's Adaptation of the TWIGL code) is a computer program for calculating the conditions in a light water reactor (LWR) core at steady state and during transients. It solves the finite-difference neutron kinetics equations on an r,z (radial, axial) mesh, the thermal-hydraulic equations for the coolant in multiple parallel, i.e., one-dimensional, channels and the one-dimensional radial fuel rod heat conduction equations for pellet, gap and clad. The analyst provides time dependent boundary conditions and/or specifications for control rod movement in order to perturb the system from an initial steady state. The boundary conditions are the inlet flow rate and temperature and a single system pressure. The analyst also supplies a normalized inlet flow distribution across the core which does not vary with time. Control rod movement includes the center rod by itself, all banks of control rods, or some combination of these. The objective of this summary is to give capabilities and limitations.
Date: January 1, 1983
Creator: Diamond, D.J. & Aronson, A.L.
Partner: UNT Libraries Government Documents Department

Reactivity transients during a blowdown in a MSIV (Main Steam Isolation Valves) closure ATWS (Anticipated Transients Without Scram)

Description: The objectives of this work are to study the consequences of the reactivity transients during a blowdown in an ATWS event with closure of the Main Steam Isolation Valves (MSIV), and to evaluate the effect of the LPCI (Low Pressure Coolant Injection) system and the sensitivity of plant response to the feedback coefficients. The present work was performed with the BNL Plant Analyzer (BPA). The BPA is a on-line, interactive BWR system code which models the non-homogeneous, non-equilibrium two-phase flow with a drift flux mixture model, the reactor kinetics with a point kinetic model, the thermal conduction with an integral method, and the control and plant protection systems with modern control theory. It also models the balance of plant (BOP) as well as the Mark I containment of a BWR/4. Thus, the BPA is a comprehensive engineering plant analyzer transients as well as accidents (e.g., ATWS and Small Break Loss of Coolant Accidents).
Date: January 1, 1988
Creator: Cheng, H.S. & Diamond, D.J.
Partner: UNT Libraries Government Documents Department

Prompt Neutron Lifetime for the NBSR Reactor

Description: In preparation for the proposed conversion of the National Institute of Standards and Technology (NIST) research reactor (NBSR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel, certain point kinetics parameters must be calculated. We report here values of the prompt neutron lifetime that have been calculated using three independent methods. All three sets of calculations demonstrate that the prompt neutron lifetime is shorter for the LEU fuel when compared to the HEU fuel and longer for the equilibrium end-of-cycle (EOC) condition when compared to the equilibrium startup (SU) condition for both the HEU and LEU fuels.
Date: June 24, 2012
Creator: Hanson, A.L. & Diamond, D.
Partner: UNT Libraries Government Documents Department

Planning the HEU to LEU Transition for the NBSR

Description: A study has been carried out to understand how the NIST research reactor (NBSR) might be converted from using high-enriched uranium (HEU) to using low-enriched uranium (LEU) fuel. An LEU fuel design had previously been determined which provides an equilibrium core with the desirable fuel cycle length - a very important parameter for maintaining the experimental, scientific program supported by the NBSR. In the present study two options for getting to the equilibrium state are considered. One option starts with the loading of an entire core of fresh fuel. This was determined to be unacceptable. The other option makes use of the current fuel management scheme wherein four fresh fuel elements are loaded at the beginning of each cycle. However, it is shown that without some alterations to the fuel cycle, none of the transition cores containing both HEU and LEU fuel have sufficient excess reactivity to operate the reactor for the optimum length. It was determined that operating the first mixed cycle for a sufficiently reduced length of time provides the excess reactivity which enables subsequent cycles to be run for the desired number of days.
Date: September 12, 2011
Creator: Hanson, A.L. & Diamond, D.
Partner: UNT Libraries Government Documents Department

Calculation of Design Parameters for an Equilibrium LEU Core in the NBSR

Description: A plan is being developed for the conversion of the NIST research reactor (NBSR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Previously, the design of the LEU fuel had been determined in order to provide the users of the NBSR with the same cycle length as exists for the current HEU fueled reactor. The fuel composition at different points within an equilibrium fuel cycle had also been determined. In the present study, neutronics parameters have been calculated for these times in the fuel cycle for both the existing HEU and the proposed LEU equilibrium cores. The results showed differences between the HEU and LEU cores that would not lead to any significant changes in the safety analysis for the converted core. In general the changes were reasonable except that the figure-of-merit for neutrons that can be used by experimentalists shows there will be a 10% reduction in performance. The calculations included kinetics parameters, reactivity coefficients, reactivity worths of control elements and abnormal configurations, and power distributions.
Date: September 30, 2011
Creator: Hanson, A.L. & Diamond, D.
Partner: UNT Libraries Government Documents Department

DETERMINATION OF INVENTORIES AND POWER DISTRIBUTIONS FOR THE NSBR.

Description: This memo presents the details of the methodology for developing fuel inventories for the NBSR along with power distributions predicted with this set of inventories. Several improvements have been made to the MCNP model of the NBSR since a set of calculations was performed in 2002 in support of the NBSR relicensing and SAR update. One of the most significant changes in the model was to divide the fuel elements into upper and lower halves so the effects of uneven burn between the two halves (due to the shim arms) can be determined. The present set of power distributions are provided for comparison with the previous safety analyses.
Date: September 12, 2005
Creator: HANSON, A.L. & DIAMOND, D.J.
Partner: UNT Libraries Government Documents Department

A Neutronics Methodology for the NIST Research Reactor Based on MCNXP

Description: A methodology for calculating inventories for the NBSR has been developed using the MCNPX computer code with the BURN option. A major advantage of the present methodology over the previous methodology, where MONTEBURNS and MCNP5 were used, is that more materials can be included in the model. The NBSR has 30 fuel elements each with a 17.8 cm (7 in) gap in the middle of the fuel. In the startup position, the shim control arms are partially inserted in the top half of the core. During the 38.5 day cycle, the shim arms are slowly removed to their withdrawn (horizontal) positions. This movement of shim arms causes asymmetries between the burnup of the fuel in the upper and lower halves and across the line of symmetry for the fuel loading. With the MONTEBURNS analyses there was a limitation to the number of materials that could be analyzed so 15 materials in the top half of the core and 15 materials in the bottom half of the core were used, and a half-core (east-west) symmetry was assumed. Since MCNPX allows more materials, this east-west symmetry was not necessary and the core was represented with 60 different materials. The methodology for developing the inventories is presented along with comparisons of neutronic parameters calculated with the previous and present sets of inventories.
Date: May 16, 2011
Creator: Hanson, A. & Diamond, D.
Partner: UNT Libraries Government Documents Department

Analyzing the rod drop accident in a BWR with high burnup fuel

Description: The response of fuel in a boiling water reactor to the rod drop accident (RDA) was studied using the RAMONA-4B computer code. Calculations of this design-basis event has been done conservatively because there was margin to the fuel failure criterion of 170 cal/g. Because high burnup fuel may fail at much lower fuel enthalpies, the best-estimate of the enthalpy and the uncertainty is of interest. In part of this study, calculations assessed the sensitivity to reactor conditions such as control rod pattern, inlet subcooling, and fuel burnup. It was shown that fuel enthalpy at any location in the region surrounding the dropped rod depends on the rod worth, the distance from the dropped rod, and the burnup of the fuel. The study also calculated the sensitivity to parameters whose modeling introduces significant uncertainty which may increase with burnup. These parameters are the control rod worth, Doppler reactivity coefficient, delayed neutron precursor fraction, and fuel specific heat. The results of the sensitivity studies were used in a model to determine the random uncertainty in the fuel enthalpy. The standard deviation for the calculated fuel enthalpy was estimated to be 37%. Therefore, the limiting bundle fuel enthalpy might be 75% higher than calculated. The effect of the fuel rod enthalpy distribution within a bundle was also investigated. RAMONA-4B calculates the fuel bundle average enthalpy and estimates must be made of a bundle peaking factor to determine the fuel rod enthalpy. A fit of RAMONA-4B bundle powers was used to estimate the local power peaking. It was determined that the peaking factor could be 25% higher than the factor usually assumed for RDA analysis. Combining this error with the random error means that for this analysis the actual fuel rod enthalpy could be 100% larger than calculated by RAMONA-4B. This is much larger ...
Date: February 1997
Creator: Diamond, D. J. & Neymotin, L.
Partner: UNT Libraries Government Documents Department

Multidimensional reactor kinetics modeling

Description: There is general agreement that for many light water reactor transient calculations, it is-necessary to use a multidimensional neutron kinetics model coupled to a thermal-hydraulics model for satisfactory results. These calculations are needed for a variety of applications for licensing safety analysis, probabilistic risk assessment (PRA), operational support, and training. The latter three applications have always required best-estimate models, but in the past applications for licensing could be satisfied with relatively simple models. By using more sophisticated best-estimate models, the consequences of these calculations are better understood, and the potential for gaining relief from restrictive operating limits increases. Hence, for all of the aforementioned applications, it is important to have the ability to do best-estimate calculations with multidimensional neutron kinetics models. coupled to sophisticated thermal-hydraulic models. Specifically, this paper reviews the status of multidimensional neutron kinetics modeling which would be used in conjunction with thermal-hydraulic models to do core dynamics calculations, either coupled to a complete NSSS representation or in isolation. In addition, the paper makes recommendations as to what should be the state-of-the-art for the next ten years. The review is an update to a previous review of the status as of ten years ago. The general requirements for a core dynamics code and the modeling available for such a code, discussed in that review, are still applicable. The emphasis in the current review is on the neutron kinetics assuming that the necessary thermal-hydraulic capability exists. In addition to discussing the basic neutron kinetics, discussion is given of related modeling (other than thermal- hydraulics). The capabilities and limitations of current computer codes are presented to understand the state-of-the-art and to help clarify the future direction of model development in this area.
Date: November 1, 1996
Creator: Diamond, D.J.
Partner: UNT Libraries Government Documents Department

Analyzing the rod drop accident in a BWR with high burnup fuel. Revised

Description: The response of fuel in a boiling water reactor to the rod drop accident (RDA) was studied using the RAMONA-4B computer code. In this study, a fit of RAMONA-4B bundle powers was used to estimate the local power peaking. It was determined that the peaking factor could be 25% higher than the factor usually assumed for RDA analysis. Combining this error with the 2 sigma random error means that for this analysis the actual fuel rod enthalpy could be 100% larger than calculated by RAMONA-4B. This is much larger than the uncertainty in most parameters that are calculated with best-estimate methods for other design-basis events.
Date: February 1, 1997
Creator: Diamond, D.J. & Neymotin, L.
Partner: UNT Libraries Government Documents Department

Analysis of a SBLOCA initiated by an ATWS event

Description: The response of a four-loop Westinghouse pressurized water reactor to SBLOCAs initiated as a result of an anticipated transient without scram (ATWS) has been analyzed using the RELAP5 computer code. The ATWS is initiated by a loss-of-feedwater, and the small breaks were due to either one or three stuck-open safety valves or reactor coolant pump seal failure. For the cases analyzed, the results show that a LOF-ATWS followed by a SBLOCA does not have more safety significance than that found when each accident is analyzed independently of one another.
Date: January 1, 1985
Creator: Pu, J.; Diamond, D.J. & Shier, W.G.
Partner: UNT Libraries Government Documents Department