Coolant Activity Research Articles

Simulation of corrosion product activity for nonlinearly rising corrosion on inner surfaces of primary coolant pipes of a typical PWR under flow rate transients

For nonlinear accelerating corrosion, calculation of activated corrosion products on inner surfaces of primary coolant pipes have been done in a typical pressurized water reactor (PWR) under flow rate perturbations. Computer program CPAIR-P (Corrosion Product Activity In Reactors) (Deeba et al., 1999) has been modified to accommodate for time-dependent corrosion rates. Results, for (24)Na, (56)Mn, (59)Fe, (58)Co, (60)Co and (99)Mo, show that the specific activity in primary loop approaches equilibrium value under normal operating conditions fairly rapidly. Predominant corrosion product activity during operation is due to (56)Mn, and cobalt isotopes dominate the activity after shutdown of reactor. Flow rate perturbations and different types of rising corrosion rates were introduced in the system and effects on saturation activity were studied. For a linear decrease in flow rate and a constant corrosion rate, the total coolant activity and activity on pipe scale approaches higher saturation values when compared to normal condition values. With a nonlinearly accelerating corrosion, the behavior of specific activity changes considerably. The flow rate perturbations on specific activity for pipe scale results in a new saturation value which depends on both the changes in flow rate (Delta w) and equilibrium corrosion rate (C(s)) values. However, the time taken to reach the saturation activity depends on the slope of corrosion rate. For a slow pump coastdown, the activity does not show an initial drop when flow rate starts decreasing. It monotonically rises and follows the slope of corrosion rate. The peak value and decay of activity after scram are strong functions of flow rate and removal efficiencies.

Nitrogen-16 Generation and Transport and Associated Shielding Requirements in a Supercritical-Water–Cooled Reactor

As a water-cooled nuclear system with a direct thermal cycle, the supercritical-water–cooled reactor (SCWR) shares with the boiling water reactor (BWR) the issue of coolant activation and transport of the coolant activation products to the turbine and balance of plant (BOP). Consistent with the BWR experience, the dominant nuclide contributing to the SCWR coolant radioactivity at full power is 16N, which is produced by an (n, p) reaction on 16O. In this paper the production and decay of 16N in the SCWR coolant circuit along with the shielding requirements imposed on the BOP are analyzed and compared with those in a BWR with a similar thermal power rating. A simple control-mass approach is adopted in which the 16N inventory in a unit mass of coolant is tracked as the coolant flows in the SCWR and BWR primary systems, which are divided into several compartments (e.g., core, lower plenum, downcomer, etc.) of known volume, mass flow rate, and neutron flux. The values of the neutron flux and (n, p) cross section in the SCWR and BWR cores are calculated by means of full-length radially reflected Monte Carlo eigenvalue models of the SCWR and BWR fuel assemblies. The results are as follows: The 16N activities in the steam lines of the BWR with normal water chemistry, in the BWR with hydrogen water chemistry, and in the SCWR are about 40, 180, and 380 μCi/g, respectively. The calculated BWR values compare well with the trends and ranges found in the literature. The SCWR 16N concentration is significantly higher than that in the BWR for the following four reasons:1. The coolant transit time in the SCWR core is about twice that in the BWR core.2. The neutron flux is higher in the SCWR because of the higher power density.3. The slow coolant pass in the water rods produces a significant 16N activity at the SCWR core inlet.4. In the SCWR all the 16N generated in the core is sent to the steam lines because there is no recirculation within the vessel.A simple gamma attenuation model shows that the higher 16N activity in the SCWR results in shielding requirements only up to 10% higher than for the BWR with hydrogen water chemistry. However, because of the higher SCWR electric power, the specific shielding costs per unit of electric power associated with 16N are expected to be similar to or better than that for BWRs.

Fuel and Fission Gas Behavior during Rise-to-Power Test of the High Temperature Engineering Test Reactor (HTTR)

The rise-to-power tests of the High Temperature Engineering Test Reactor (HTTR) have been carried out successfully by the Japan Atomic Energy Research Institute (JAERI). For the safe operation of HTTR, the continuous and reliable measurement of the coolant activity is required to evaluate the fuel performance during normal operating conditions. In order to measure the primary coolant radioactivity (PCR), the PCR instrumentation of the safety protection system, the fuel failure detection (FFD) system and the primary coolant sampling system have been installed in the primary circuit. The PCR was less than 103 MBq/m3, and measured Kr and Xe isotopes were less than 0.1 MBq/m3 during the rise-to-power tests. The measured fractional releases are constant at 2x10-9 up to 60% of the reactor power, and then increase to 7x10-9 at full power operation. The prediction shows good agreement with the measured value. These results showed that the release mechanism varied from recoil to diffusion of the generated fission gas from the contaminated uranium in the fuel compact matrix.

Computer simulation of corrosion product activity in primary coolants of a typical PWR under flow rate transients and linearly accelerating corrosion

Computer simulation of behavior of coolant activation due to corrosion products have been investigated in a typical pressurized water reactor (PWR) under flow rate perturbations for linearly accelerating corrosion. The computer program CPAIR-P ( Deeba et al., 1999) has been modified to accommodate for time dependent corrosion. Results for 24Na, 56Mn, 59Fe, 60Co and 99Mo show that the specific activity in primary loop approaches equilibrium value under normal operating conditions fairly rapidly. During reactor operation, predominant corrosion product activity is due to 56Mn and after shutdown cobalt activity dominates. These simulations suggest that the effect of flow rate perturbations on specific activity in the form of a depression in the activity curve can be smeared by a linearly rising corrosion. Such a dip can only be seen in activity when corrosion rate approaches to an equilibrium value well before the initiation of the transient. The time period to reach minimum coolant activity during transient is a function of the slope of flow rate perturbation parameter, g( t). The new saturation value for activity depends on changes in flow rate (Δ w) and equilibrium value ( C s) for the corrosion rate. For linearly accelerated corrosion and a pump coastdown condition, the activity does not show an initial drop when flow starts decreasing. It monotonically rises and follows the slope of corrosion rate. If the slope is greater than 9×10 −6 μg/s 2, then the activity crosses the normal saturation value of 0.22 μCi/cm 3 before the reactor scram occurs. These results also indicate that the pump coastdown does produce a coolant activity spike before the reactor scram; however it can only be observed when the corrosion acceleration is fast.

Modelling the activity of 129I in the primary coolant of a CANDU reactor

A mathematical treatment has been developed to describe the activity levels of 129I as a function of time in the primary heat transport system during constant power operation and for a reactor shutdown situation. The model accounts for a release of fission-product iodine from defective fuel rods and tramp uranium contamination on in-core surfaces. The physical transport constants of the model are derived from a coolant activity analysis of the short-lived radioiodine species. An estimate of 3×10 −9 has been determined for the coolant activity ratio of 129I/ 131I in a CANDU Nuclear Generating Station (NGS), which is in reasonable agreement with that observed in the primary coolant and for plant test resin columns from pressurized and boiling water reactor plants. The model has been further applied to a CANDU NGS, by fitting it to the observed short-lived iodine and long-lived cesium data, to yield a coolant activity ratio of ∼2×10 −8 for 129I/ 137Cs. This ratio can be used to estimate the levels of 129I in reactor waste based on a measurement of the activity of 137Cs.

Prediction of Fuel Performance and Fission Gas Release Behavior during Normal Operation of the High Temperature Engineering Test Reactor by JAERI and FZJ Modeling Approach

In high temperature gas-cooled reactors (HTGRs), coated fuel particles are employed as fuel to permit high outlet coolant temperature. The essential feature of HTGRs is the role of the coated fuel particles acting as tiny containment and the principal barrier against radionuclide release under any operational and accident condition. For a safe operation of the High Temperature Engineering Test Reactor (HTTR), the continuous and reliable measurement of the coolant activity is required to allow an evaluation of the fuel performance and the radiological assessment of the plant during normal operation conditions. Since fission gases do not plateout on the inner surface of the primary cooling system, their concentrations in the primary coolant reflect the core-average through-coatings failure fraction and the fuel matrix contamination fraction. The main purpose of this report was to give a prediction of the fuel performance and fission product behavior in the HTTR under normal operating conditions by applying the calculation models as used at Japan Atomic Energy Research Institute and Research Center Juelich (FZJ), Germany, and to compare the results and methodologies. In the prediction of core average failure fraction, JAERI model gave much earlier failure and about two times larger than FZJ model. In the fission gas release prediction, FZJ model showed later increase of fractional release than JAERI model and rapidly increased towards the end of the operation.

Occupational doses and impact on fusion economics

As part of the SEAFP-99 programme, water cooled and helium cooled fusion power plant design concepts were assessed with regard to occupational doses and related availability and operating costs. Different design and management measures to reduce occupational doses were considered. This task assessed the existing designs and proposed extensions or changes, where required. For each significant contributor, possible methods were specified which could reduce the operator dose and outage time. Where this was the case expected; costs or savings incurred by the improvement were estimated. Overall, the use of a system to remove particulate corrosion product activity from the primary coolant was considered essential for the water-cooled variant. In addition, application of the most up-to-date ALARA techniques could make the estimated dose for the water-cooled conceptual design appreciably lower than earlier estimated. It was concluded that the water cooled conceptual design could meet the project design target for occupational dose of 0.7 p-Sv per GW(e). A survey of occupational doses from the UK AGR plants was also carried out. From this, it is judged that the helium cooled fusion plant conceptual design could meet the project design target for occupational dose of 0.7 p-Sv per GW(e) without the need for additional plant to reduce primary coolant activity levels.

Prediction of Fuel Performance and Fission Gas Release Behavior during Normal Operation of the High Temperature Engineering Test Reactor by JAERI and FZJ Modeling Approach.

In high temperature gas-cooled reactors (HTGRs), coated fuel particles are employed as fuel to permit high outlet coolant temperature. The essential feature of HTGRs is the role of the coated fuel particles acting as tiny containment and the principal barrier against radionuclide release under any operational and accident condition. For a safe operation of the High Temperature Engineering Test Reactor (HTTR), the continuous and reliable measurement of the coolant activity is required to allow an evaluation of the fuel performance and the radiological assessment of the plant during normal operation conditions. Since fission gases do not plateout on the inner surface of the primary cooling system, their concentrations in the primary coolant reflect the core-average through-coatings failure fraction and the fuel matrix contamination fraction. The main purpose of this report was to give a prediction of the fuel performance and fission product behavior in the HTTR under normal operating conditions by applying the calculation models as used at Japan Atomic Energy Research Institute and Research Center Juelich (FZJ), Germany, and to compare the results and methodologies. In the prediction of core average failure fraction, JAERI model gave much earlier failure and about two times larger than FZJ model. In the fission gas release prediction, FZJ model showed later increase of fractional release than JAERI model and rapidly increased towards the end of the operation.

Copper corrosion and activation in water cooling loops under fusion irradiation conditions

Copper alloys have recently been proposed in the design of the International Thermonuclear Experimental Reactor (ITER), Spherical Tokamak (ST) power plants (ARIES ST, Culham ST), and have been used in the construction of the central column of experimental machines like MAST (UKAEA) and NSTX (PPPL). The activation–corrosion code transport of activation (TRACT) is used to predict corrosion, transport behaviour, and activity in the central column-cooling loop of the ST power plant. A 2 yr irradiation of the central column results in coolant activity due to the 66Cu, 62Cu, 16N and 64Cu nuclides. After power down, activity is dominated by 64Cu, but after a few days it is due to the long-lived nuclides. The solubility and radiolysis models developed here can be used to predict corrosion release rates and the necessary water chemistry conditions to minimise activated products in cooling loops of fusion experimental machines and future power plants.

Modeling and simulation of corrosion product activity in pressurized water reactors under power perturbations

An improved methodology is presented for simulation of coolant activation due to corrosion products and impurities in a typical pressurized water reactor (PWR) under power perturbations. Using time dependent production and losses of corrosion products in the primary coolant path an approach has been developed to calculate the coolant specific activity. Results for 24Na, 56Mn, 59Fe, 60Co and 99Mo show that the specific activity in primary loop approaches equilibrium value under normal operating conditions fairly rapidly. Predominant corrosion product activity is due to 56Mn. Flow rate has been assumed to follow power changes and different types of power perturbations are introduced after the equilibrium activity has been achieved. In particular the effects of linear changes in reactor operating power and power peaking on the corrosion product activity of the primary coolant have been studied.

98/02951 Development of a computer code to estimate the fuel rod failure using primary coolant activities of operating PWRs

98/02951 Development of a computer code to estimate the fuel rod failure using primary coolant activities of operating PWRs

Development of a computer code to estimate the fuel rod failure using primary coolant activities of operating PWRS

A Windows computer code entitled as “CAAP (Cooland Activity Analysis Program)” has been developed to evaluate the number, the degree of failures, and the location of failed fuel rods using primary coolant radioactivity data obtained from operating PWRS. New models are developed and incorporated into the CAAP program to improve some of the drawbacks of existing computer codes. The iodine and noble gas activities obtained by grab sampling and that obtained from the primary coolant radioactivity monitoring system can be used to estimate fuel rod failures with CAAP. For an on-line evaluation of the number of failed fuel rods and the degree of failures, CAAP has been combined with the primary coolant radioactivity monitoring system at Kori 3 & 4 operating nuclear power plants. The validity of the computational models of CAAP has been examined using nuclear power plant data collected. For 29 cycles of PWRs for which ultrasonic inspections were performed at the end of fuel cycles, the number of failed fuel rods has been estimated with CAAP and compared with the ultrasonic inspection results. The results show that CAAP gives a better agreement with the ultrasonic test data than existing models. In addition, CAAP calculations performed to estimate the region and the burnup of the failed fuel rods for 2 cycles, whose locations of failed fuel rods were known, show that CAAP can predict both the region and the burnup of the failed fuel rods accurately.

Texaco acquires Dow's German coolant business activity

Texaco acquires Dow's German coolant business activity

Simulation of corrosion product activity in pressurized water reactors under flow rate transients

Simulation of coolant activation due to corrosion products and impurities in a typical pressurized water reactor has been done under flow rate transients. Employing time dependent production and losses of corrosion products in the primary coolant path an approach has been developed to calculate the coolant specific activity. Results for 24Na, 56Mn, 59Fe, 60Co and 99Mo show that the specific activity in primary loop approaches equilibrium value under normal operating conditions fairly rapidly. Predominant corrosion product activity is due to Mn-56. Parametric studies at full power for various ramp decreases in flow rate show initial decline in the activity and then a gradual rise to relatively higher saturation values. The minimum value and the time taken to reach the minima are strong functions of the slope of linear decrease in flow rate. In the second part flow rate coastdown was allowed to occur at different flow half-times. The reactor scram was initiated at 90% of the normal flow rate. The results show that the specific activity decreases and the rate of decrease depends on pump half time and the reactor scram conditions.

Neutron activation of lithium coolant in the V5Ti5Cr alloy circuit

In the current phase of ITER design activity, a possibility to use a lithium-cooled test module with water-cooled FW/blanket and vacuum vessel at the BPP is discussed. In this paper neutron activation of the lithium coolant, circulating in the ITER V5Ti5Cr alloy circuit and in the stainless steel circuit, is calculated. The results have shown, that in latter case the total long-lived γ-activity of coolant is greater by an order of two.

Effect of flow rate and power perturbations on dose rates due to coolant activity in low-power research reactors

A computational approach has been developed using the multigroup diffusion theory based leopard and odmug codes to study the dose rate due to coolant activation in typical low-power tank-in-pool-type research reactors. A one-dimensional coolant flow model is used for natural circulation to calculate the activity reaching the upper layers of water in the reactor vessel. The effect of the flow rate and power perturbations on dose rates at two different locations in the reactor building has been determined. The results show that, with a decrease in the flow rate, the activity leaving the core increases. The total dose rate at the reactor bridge (about 6 m above the core) changes from 60.1 to 27.5 μSv/h as the flow rate is decreased from 0.36 to 0.1 kg/s in the core. For steady-state power, the difference between the experimental and theoretical dose rate values remains < 8%. In a power excursion experiment when 4 mk reactivity was released by fully withdrawing the control rod, the fast and thermal fluxes increased by a factor of 2.7 as the power rose from 30 to 87 kW. The dose rate at the reactor bridge due to all isotopes initially increases to a maximum value as both the flow rate and power rise. In the initial 300 s, the rise in both the flow rate and power cause the total dose rate at the reactor bridge to reach a maximum of about 1.51 mSv/h. The dose rate lags behind the power excursion and reaches a second peak (∼ 1.33 mSv/h) at 390 s due to power peaking at about 350 s after the initiation of the transient. The major source of the total dose rate is due to 16N and other isotopes contribute < 0.2 mSv/h during the transient.

A Prototype Expert System for the Monitoring of Defected Nuclear Fuel Elements in Canada Deuterium Uranium Reactors

This paper reports on a prototype expert system for fuel failure monitoring in Canada deuterium uranium (CANDU) power reactors. Based on a coolant activity analysis, the system is able to provide information in an operating reactor on the number of fuel failures, the average defect size, and the amount of tramp uranium deposited on the in-core surfaces of the primary heat transport system. The fission product release model used in the system is based on results from an in-reactor experimental program at Chalk River Nuclear Laboratories. The expert system is validated against fuel failure data from a number of CANDU power reactors.

Study of Coolant Activation and Dose Rates with Flow Rate and Power Perturbations in Pool-Type Research Reactors

This paper reports on a computer code using the multigroup diffusion theory based LEOPARD and ODMUG programs that has been developed to calculate the activity in the coolant leaving the core of a pool-type research reactor. Using this code, the dose rates at various locations along the coolant path with varying coolant flow rate and reactor power perturbations are determined. A flow rate decrease from 1000 to 145 m{sup 3}/h is considered. The results indicate that a flow rate decrease leads to an increase in the coolant outlet temperature, which affects the neutron group constants and hence the group fluxes. The activity in the coolant leaving the core increases with flow rate decrease. However, at the inlet of the holdup tank, the total dose rate first increases, then passes through a maximum at {approximately} 500 m{sup 3}/h, and finally decreases with flow rate decrease. The activity at the outlet of the holdup tank is mainly due to {sup 24}Na and {sup 56}Mn, and it increases by {approximately} 2% when the flow rate decreases from 1000 to 145 m{sup 3}/h. In an accidental power rise at constant flow rate, the activity in the coolant increases, and the dose rates at allmore » the points along the coolant path show a slight nonlinear rise as the reactor power density increases.« less

Nuclear Power Reactor Development in Japan

Based on the current situation in Japan that nearly 40 LWRs (PWR/BWR nearly 50/50) have been operating with excellent performance; the number of un-intended shutdown is the smallest, leaker fuel is practically zero and thus coolant activity is the lowest in the world, they proceed to introduce advanced LWR (ALWR) of 1,300 MWe in early '90s, followed by the further improvement into higher duty LWR (so-called next generation LWR). National program for advanced thermal reactor (ATR) and FBR development has been steadily performed; a demonstration reactor of ATR and a prototype of FBR are under construction, which are to be put into operation by middle 1990s. Lately-revised high temperature gas cooled reactor program for process that generation is put into practice in 1989 and its experimental reactor to be used also for irradiation facility will be completed 7 years later. Smaller-size LWR and so-called «inherent safety type reactor» have also been interested. Overview on those activities are presented

Defective Fuel Rod Detection in Operating Pressurized Water Reactors During Periods of Continuously Decreasing Fuel Rod Integrity Levels

Periods of continuously decreasing levels of fuel rod integrity due to debris-induced cladding damage, vibration-induced fretting wear of the cladding, etc., cause difficulties in the assessment of fuel rod performance from coolant activity data. The calculational models currently in use for this purpose in nuclear power plants are not sufficiently capable of indicating cases in which they are invalid. This can mislead reactor operators by misinterpretation of the coolant activity data, especially in situations where fast reactions are necessary. A quick test of validity is suggested to check the applicability of the currently available calculational models for estimating the number and average size of fuel rod defects. The criteria of applicability are the fission product total activity, the slope tanα in the relationship Ig(Ri/Bi) = tanαlg λi for the volatile fission product isotopes, and the activity of nonvolatile nuclides such as 239Np and the isotopes of ruthenium and cerium. The objective is to recognize immediately periods of continuously decreasing levels of fuel rod integrity in order to prevent complications in routine power plant maintenance as well as accident situations caused by more severe fuel rod degradation.