Peach Bottom Unit Research Articles

Validation of the SCALE/Polaris–PARCS Code Procedure With the ENDF/B-VII.1 AMPX 56-Group Library: Boiling Water Reactor

The SCALE/Polaris–PARCS code procedure has been used in the confirmatory analysis for boiling water reactors by the US Nuclear Regulatory Commission. In this study, the SCALE/Polaris v6.3.0–PARCS v3.4.2 code procedure with the Evaluated Nuclear Data File (ENDF)/B-VII.1 AMPX 56-group library was validated by comparing the simulated results with the measured data for operating boiling water reactors, including Peach Bottom Unit 2 cycles 1–3, Hatch Unit 1 cycles 1–3, and Quad Cities Unit 1 cycles 1–3. The uncertainties and biases of the SCALE/Polaris–PARCS code package for boiling water reactor physics analysis were evaluated in the validation for key nuclear parameters such as reactivity and traversing in-core probe data.

McCARD/MIG stochastic sampling calculations for nuclear cross section sensitivity and uncertainty analysis

In this study, a cross section stochastic sampling (S.S.) capability is implemented into both the McCARD continuous energy Monte Carlo code and MIG multiple-correlated data sampling code. The ENDF/B-VII.1 covariance data based 30 group cross section sets and the SCALE6 covariance data based 44 group cross section sets are sampled by the MIG code. Through various uncertainty quantification (UQ) benchmark calculations, the McCARD/MIG results are verified to be consistent with the McCARD stand-alone sensitivity/uncertainty (S/U) results and the XSUSA S.S. results. UQ analyses for Three Mile Island Unit 1, Peach Bottom Unit 2, and Kozloduy-6 fuel pin problems are conducted to provide the uncertainties of keff and microscopic and macroscopic cross sections by the McCARD/MIG code system. Moreover, the SNU S/U formulations for uncertainty propagation in a MC depletion analysis are validated through a comparison with the McCARD/MIG S.S. results for the UAM Exercise I-1b burnup benchmark. It is therefore concluded that the SNU formulation based on the S/U method has the capability to accurately estimate the uncertainty propagation in a MC depletion analysis.

The status quo on HTGR decommissioning

Five High-Temperature Gas-Cooled Reactors, which have been operated in the past, are now in different stages of decommissioning. The present status, technical approaches and experiences are described for AVR, DRAGON, FSV, Peach Bottom Unit 1 and THTR.After defueling the core, the Reactor Pressure Vessel (RPV) of AVR has been grouted with low-density cellular concrete and removed from the containment building. The RPV is now stored in a new interim storage building, on the site of the Juelich research center. The DRAGON reactor was operated in UK from 1965 until 1976 when it was defueled and put into safe storage. It was partially decommissioned in 2005 and is now facing further removal steps with a target decommissioning completion date of 2022. The 330 MWel Fort Saint Vrain (FSV) HTGR in US was shut down in 1989 and undergone immediate dismantling with a fixed-price contractor. The plant site was released for unrestricted use in August 1997. The Peach Bottom Unit 1 was shut down in 1974. The spent fuel was transferred to INL. Decommissioning is not expected until 2034, when the other LWR units at this site are scheduled to shut down. The German 300 MWel Thorium High-Temperature Reactor (THTR) was closed-down in 1989 and resides under Safe-Enclosure conditions since 1997, after having evaluated the radiological consequences of direct dismantling vs. 30 years safe enclosure.The 32 MWel Windscale Advanced Gas-Cooled Reactor (WAGR) prototype shows some relevant design similarities to modular HTGR. Lessons learned from the WAGR and other relevant decommissioning projects are therefore addressed, too.As a general conclusion, it can be stated that the decommissioning procedures need to be thoroughly reflected already during the design of any new HTGR project.The paper has been elaborated within the European GEMINI+ project and is based on the deliverables of the Task ‘Decommissioning and Waste Management of the GEMINI+ system’.

Turbopump scaling analysis and similarity level estimation for Texas A&M university RCIC system experimental test facility

The Reactor Core Isolation Cooling (RCIC) System is a safety system that provides water to the reactor pressure vessel during off-normal BWR reactor conditions, such as reactor isolation from the turbines or loss of AC power. Under loss of AC power conditions, the RCIC System is expected to fail due to battery depletion within 4–8 h of operation for many units. However, the system did not fail until about 70 h into the accident at Fukushima Dai-ichi Unit 2, which was well past the time of battery depletion. To investigate the full potential of the RCIC System, the Laboratory for Nuclear Heat Transfer Systems (NHTS) at Texas A&M University is designing and constructing an experimental RCIC System facility. A careful scaling analysis is essential to ensure proper representation of the RCIC System's key components and phenomena in the experimental testing.This paper describes a method to estimate the similarity level (SL) of the RCIC turbine-pump assembly (turbopump), along with application of the methodology using the Peach Bottom Unit 2 RCIC System as the reference full-scale system. The methodology is demonstrated with the Texas A&M University facility but can be applied to other RCIC system facilities.Zuber's H2TS (Hierarchal Two-Tiered Scaling) method has been adopted to specify the required level of detail for the scaling analysis. The analysis includes development of unitless characteristic time ratios to quantify the degree of similarity of the RCIC turbomachinery between the test facility and the prototype. The RCIC turbine and pump are focused on herein because, first, the Fukushima Dai-ichi Unit 2 RCIC System may have ceased operation at 70 h due to turbine failure, and, two, the greatest RCIC System modeling challenges are with the long-term operation of these turbomachinery components. Since the Texas A&M experiments will provide validation data for analytical models of the turbomachinery, applicability of the data must be evaluated.Based on the similarity level estimates, there is a high degree of similarity between the NHTS and Peach Bottom RCIC System turbines, with a minimum similarity level of 75%. This allows the use of the NHTS RCIC turbine for testing over the full range of Design Basis Accident conditions. Another important finding from this study is that the pump configuration initially designed into the NHTS facility had a low similarity level between the experimental and full-scale facilities. This paper identifies pump parameters in the NHTS facility that can be changed to achieve greater similarity. The similarity level analysis indicate that the NHTS RCIC turbine control volume is appropriate for representing the full-size turbomachinery and can be used to study the prototype turbine and pump behaviors under design basis operating conditions.Scaling analyses of the other RCIC System components are to follow the turbomachinery similarity evaluations. The NHTS facility is currently being prepared to test over the entire Design Basis Accident containment pressure and temperature ranges. Future efforts will also focus on modeling, experimentally and with Computational Fluid Dynamics tools, the RCIC System to provide data for beyond design basis operational conditions.

An optimally diffusive Coarse Mesh Finite Difference method to accelerate neutron transport calculations

A new optimally diffusive Coarse Mesh Finite Difference (odCMFD) method is proposed. The new method generalizes the Coarse Mesh Finite Difference (CMFD) and partial current-based CMFD (pCMFD) methods by adding an artificial term to the diffusion coefficient. The standard CMFD and pCMFD methods preserve the net current and partial currents respectively and can greatly reduce the spectral radius for solving neutron transport problems. Linearized Fourier analysis shows that these methods essentially differ only by the definition of the diffusion coefficient – the pCMFD diffusion coefficient contains an additional “artificial” term, usually taken to be ¼ the coarse mesh size. In this paper, the magnitude of the artificially diffusive term is numerically investigated and optimized using Fourier analysis. The results show that the optimal coefficient increases as the coarse cell optical thickness increases. Also, the optimal value is always smaller than the pCMFD value. This indicates that the pCMFD method overcorrects the diffusion coefficient which increases the spectral radius. A simple polynomial fitting was found for implementing the odCMFD in the MPACT code to perform calculations for realistic problems. Numerical MPACT results for a two-region homogeneous problem, a 2D C5G7 problem, a VERA problem 5 and a 2D BWR Peach Bottom Unit 2 core problem agree well with the Fourier analysis and confirm previous research results that: (1) CMFD converges faster than pCMFD for optically thin coarse grids, but CMFD diverges for optically thick coarse grids; (2) pCMFD is unconditionally stable for all coarse grids; and show that the newly proposed odCMFD is the most efficient among these three methods for all cases. The new method has a similar convergence rate as CMFD for optically thin coarse grids, is faster than both CMFD and pCMFD for coarse grids with intermediate optical thickness, and is unconditionally stable and faster than pCMFD for optically thick coarse grids.

OECD/NEA BENCHMARK FOR UNCERTAINTY ANALYSIS IN MODELING (UAM) FOR LWRS – SUMMARY AND DISCUSSION OF NEUTRONICS CASES (PHASE I)

A Nuclear Energy Agency (NEA), Organization for Economic Co-operation and Development (OECD) benchmark for Uncertainty Analysis in Modeling (UAM) is defined in order to facilitate the development and validation of available uncertainty analysis and sensitivity analysis methods for best-estimate Light water Reactor (LWR) design and safety calculations. The benchmark has been named the OECD/NEA UAM-LWR benchmark, and has been divided into three phases each of which focuses on a different portion of the uncertainty propagation in LWR multi-physics and multi-scale analysis. Several different reactor cases are modeled at various phases of a reactor calculation. This paper discusses Phase I, known as the “Neutronics Phase”, which is devoted mostly to the propagation of nuclear data (cross-section) uncertainty throughout steady-state stand-alone neutronics core calculations. Three reactor systems (for which design, operation and measured data are available) are rigorously studied in this benchmark: Peach Bottom Unit 2 BWR, Three Mile Island Unit 1 PWR, and VVER-1000 Kozloduy-6/Kalinin-3. Additional measured data is analyzed such as the KRITZ LEU criticality experiments and the SNEAK-7A and 7B experiments of the Karlsruhe Fast Critical Facility. Analyzed results include the top five neutronnuclide reactions, which contribute the most to the prediction uncertainty in keff, as well as the uncertainty in key parameters of neutronics analysis such as microscopic and macroscopic cross-sections, six-group decay constants, assembly discontinuity factors, and axial and radial core power distributions. Conclusions are drawn regarding where further studies should be done to reduce uncertainties in key nuclide reaction uncertainties (i.e.: 238U radiative capture and inelastic scattering (n, n’) as well as the average number of neutrons released per fission event of 239Pu).

Analyses of pressure perturbation events in boiling water reactor

In order to design more stable and safer core configurations, experimental and theoretical studies about BWR (Boiling Water Reactor) instability have been performed to characterize the phenomenon and to predict the conditions for its occurrence. The instabilities can be caused by interdependencies between thermal-hydraulic and reactivity feedback parameters such as the void-coefficient, for example, during a pressure perturbation event. In this work, the RELAP5-MOD3.3 thermal-hydraulic system code and the PARCS-2.4 3D neutron kinetic code were coupled to simulate BWR transients. The pressure perturbation is considered in order to study in detail this type of transient. Two different algorithms developed at the University of Pisa were used to calculate the Decay Ratio (DR) and the natural frequency (NF) from the power oscillation signals obtained from the transient calculations. The validation of a code model set up for the Peach Bottom-2 BWR plant is performed against Low-Flow Stability Tests (LFST). The four series of Stability Tests were performed at Peach Bottom Unit 2 in 1977 at the end of cycle 2 in order to measure the reactor core stability margins at the limiting conditions used in design and safety analysis.

Analyses of Instability Events in the Peach Bottom-2 BWR Using Thermal-Hydraulic and 3D Neutron Kinetic Coupled Codes Technique

Boiling water reactor (BWR) instabilities may occur when, starting from a stable operating condition, changes in system parameters bring the reactor towards an unstable region. In order to design more stable and safer core configurations, experimental and theoretical studies about BWR stability have been performed to characterise the phenomenon and to predict the conditions for its occurrence. In this work, contributions to the study of BWR instability phenomena are presented. The RELAP5/MOD3.3 thermal-hydraulic (TH) system code and the PARCS-2.4 3D neutron kinetic (NK) code were coupled to simulate BWR transients. Different algorithms were used to calculate the decay ratio (DR) and the natural frequency (NF) from the power oscillation predicted by the transient calculations as two typical parameters used to provide a quantitative description of instabilities. The validation of the code model set up for the Peach Bottom Unit 2 BWR plant is performed against low-flow stability tests (LFSTs). The four series of LFST have been performed during the first quarter of 1977 at the end of cycle 2 in Pennsylvania. The tests were intended to measure the reactor core stability margins at the limiting conditions used in design and safety analyses.

Analysis of the OECD/NRC BWR Turbine Trip Transient Benchmark with the Coupled Thermal-Hydraulics and Neutronics Code TRAC-M/PARCS

An analysis of the Peach Bottom Unit 2 Turbine Trip 2 (TT2) experiment has been performed using the U.S. Nuclear Regulatory Commission coupled thermal-hydraulics and neutronics code TRAC-M/PARCS. The objective of the analysis was to assess the performance of TRAC-M/PARCS on a BWR transient with significance in two-phase flow and spatial variations of the neutron flux. TRAC-M/PARCS results are found to be in good agreement with measured plant data for both steady-state and transient phases of the benchmark. Additional analyses of four fictitious extreme scenarios are performed to provide a basis for code-to-code comparisons and comprehensive testing of the thermal-hydraulics/neutronics coupling. The obtained results of sensitivity studies on the effect of direct moderator heating on transient simulation indicate the importance of this modeling aspect.

RETRAN-3D MOD003 Peach Bottom Turbine Trip 2 Multidimensional Kinetics Analysis Models and Results

An analysis of the Peach Bottom Unit 2 Turbine Trip Test 2 (PB2/TT2) has been performed using RETRAN-3D MOD003. The purpose of the analysis was to investigate the PB2/TT2 overpressurization transient using the RETRAN-3D multidimensional kinetics model.

Analytical Study on Detailed Void Distributions Inside BWR Fuel Bundle under Turbine Trip Event Considering Time-Dependent Pin Power Distributions

A one-way coupling system between the plant simulator TRAC/BF1-ENTRÉE and the subchannel code with the improved cross flow model, NASCA, has been developed. Based on a scenario of turbine trip tests in the Peach Bottom Unit 2, the wide and rapid reactivity insertion transient induced by the system pressure rise was calculated. The pin power distribution in hot bundles was re-constructed considering heterogeneity of the fuel bundle. When the neighboring control blade is withdrawn, NASCA predicted that the steady-state bundle exit void distribution was nearly flat with regardless of the pin power distribution. However, void distributions in the middle and lower bundle regions became complicated depending on the pin power and the two-phase flow regime in each subchannel. The pin power distribution rapidly changed according to traveling of control blades. However, influence in the void distribution was delayed and damped due to the fuel heat conduction. The detailed void distribution under transient events has an impact on the location of dryout. It has been shown that the coupling calculation including the subchannel code is useful in understanding transition of the detailed void distribution depending on the pin power and the two-phase flow regime.

Benchmarking and Qualification of Retran-03 for BWR Stability Analysis by Comparison with Frequency-Domain Stability Analysis Code

The benchmarking and qualification analyses of RETRAN-03 (RETRAN-3D) for boiling water reactor (BWR) stability analyses were carried out by comparison with the frequency-domain stability analysis code NUFREQ-NPT with the stability test data of the Peach Bottom Unit 2. The sensitivities of model parameters were studied in terms of the type of equation model, vapor-liquid interface heat transfer coefficient in upper downcomer, method of characteristics (MOC) model, proportionality constant in the pressure change mass transfer term, and nodalization of a core for the turbine trip test analyses. The sensitivity studies of the model parameters to the decay ratio in stability analyses were performed on the number of core channels, type of equation model, nodalization of a core, perturbation type of disturbance, slip model, proportionality constant in the pressure change mass transfer term, Courant number, MOC model, and kinetics model. The models were selected for the turbine trip tests analyses and for stability tests analyses, based on the sensitivity studies. The model used to analyze stability in RETRAN-03 adopted the five-equations with the MOC, and two-channel models for the core heating region divided into 40 nodes despite 24 nodes used for the turbine trip test analyses. The validation of the model was confirmed by the analyses of the turbine trip tests of the Peach Bottom Unit-2. The stability analyses with the test data and the benchmarking of RETRAN-03 compared with the frequency-domain stability analysis code NUFREQ-NPT in BWR stability exhibit verification and validation within the applicable limitation of the code.

Thermal-Hydraulic Analysis of a Hypothetical Complete Failure of One Shroud Access Hole in a Boiling Water Reactor

Cracks have been observed in the welds of the shroud support access hole cover plates at Peach Bottom Unit 3 as well as at the Chinshan nuclear power plant. The cracks could potentially result in complete weld failure and, as a consequence, a coreflow bypass. Analysis of the postulated complete weld failure of the Chinshan nuclear power plant shroud support access hole cover plate is performed with RETRAN-02/MOD3. The results show that some confusing phenomena could be observed. The reactor operator should be aware of the phenomena to avoid the likelihood of misoperation. Also, a special procedure needs to be established to deal with this event.

Evaluation of RETRAN-02 Options with Peach Bottom Data

The RETRAN-01 system thermal-hydraulics analysis code has been used to simulate the Peach Bottom Unit 2 boiling water reactor (BWR) turbine trip tests. Since these studies have been performed, a second version of the code, RETRAN-02, has been developed and released. The RETRAN-02 code has a number of physical models that did not exist in RETRAN-01. These include slip between phases, subcooled voiding, special steam separator, one-dimensional kinetics, and iterative numerics models. The new models that exist in RETRAN-02 allow a more realistic modeling of the physical phenomena that occur, in particular, in a BWR. The effect of these new models is studied through the analysis of one of the Peach Bottom turbine trip tests. Each model's effect is easily observed since the models are applied one at a time. It was demonstrated that if the same options are used in both RETRAN-01 and RETRAN-02, essentially the same results are obtained. This is important because of the large amount of validation that has been performed with RETRAN-01. The results using the new models demonstrate that they are capable of more realistically describing the physical phenomena that occur during a turbine trip in a BWR.

Steam Generator Materials Performance in High Temperature Gas-Cooled Reactors

High temperature gas-cooled reactor (HTGR) systems feature a graphite-moderated, uranium-thorium, all-ceramic core and utilize high pressure helium as the primary coolant. The steam generators in these systems are exposed to gas-side temperatures approaching 760°C (1400°F) and produce superheated steam at 538°C (1000°F) and 16.5 MPa (2400 psi). Thus, the design and development of steam generators for these systems require consideration of time-dependent materials behavior, corrosion, fretting, wear, and other related phenomena of concern in all steam generators.The prototype Peach Bottom Unit No. 1 40-MW (electric) HTGR was operated by the Philadelphia Electric Company for a total of 1349 equivalent full power days during a 7-yr period. Upon planned decommissioning of that plant, the forced-recirculation U-tube steam generators and other components were subjected to destructive properties tests and metallurgical examinations. These tests and examinations showed the steam generators to be in very satisfactory condition.The 330-MW(electric) Fort St. Vrain HTGR, owned and operated by Public Service Company of Colorado, and now in the final stages of startup, has achieved 70% power and generated more than 1.5 × 106 MWh of electricity. The steam generators in this reactor are once-through units of helical configuration, and their design and development required considering a number of new materials factors including creep fatigue. Also, because of the once-through design, water chemistry control needed special consideration.Current designs of larger HTGRs also feature steam generators of helical tube once-through design. Materials issues that are important in these designs include detailed consideration of time-dependent behavior of both base metals and welds, as required by current American Society of Mechanical Engineers Code rules, evaluation of bimetallic weld behavior, evaluation of the properties of very large tubesheet forgings, consideration of the gaseous corrosion effects of the primary coolant, and other related factors.

RETRAN Analysis of the Peach Bottom Turbine Trip Tests

The RETRAN system thermal-hydraulics analysis code has been used to model a series of three turbine trip tests. These three tests were performed at the Peach Bottom Unit 2 (PB2) boiling water reactor (BWR) during April 1977. A detailed description of this analysis including sensitivity studies using RETRAN-01 has been issued by the Electric Power Research Institute (EPRI). Additional analysis using RETRAN-02 is described. 8 refs.

Peach bottom unit no. 1: A high performance helium cooled nuclear power plant

The first prototype high temperature gas cooled reactor (HTGR) was successfully developed, designed, built, tested, operated and decommissioned between 1958 and 1978 as part of a cooperative program between the Atomic Energy Commission and several industries participating as follows: Philadelphia Electric Company, owner-operator together with High Temperature Reactor Development Associates; General Atomic Company, conceptual development design and NSSS supplier; and Bechtel Corporation, architect engineer. Developed improvements included: the successful replacement of austenitic stainless steel steam generator tubes that suffered chloride stress corrosion with Incoloy tubes; and the replacement of purged fuel elements, consisting of uranium and thorium carbide particles coated with a dense pyrocarbon in a graphite matrix, with irradiation resistant double coated particles that retained fission products and yielded no more than 0.5 Ci radioactivity in the continuously purified helium coolant. In commercial operation, the plant produced steam at 1000F and 1450 psig and generated power with a thermal efficiency of 37% and 88% availability, excluding R and D programs and planned shutdowns, from July 1967 to October 1974. Termination of operation and decommissioning proceeded according to approved plans.