Primary Coolant Loop Research Articles

Application of LBB to high energy pipings of a pressurized water reactor in Korea

The application of the leak before break (LBB) technology to the newly constructed pressurized water reactors (PWRs) has been approved in Korea for several high energy systems that can meet rigorous acceptance criteria. The LBB application in Korea is based on the US Nuclear Regulatory Commision (USNRC) requirements. The purpose of the LBB evaluation is to eliminate the dynamic effects associated with the postulated double-ended guillotine break (DEGB) from design basis loads, as well as to eliminate pipe whip restraints and jet impingement barriers. There were several issues on the application of LBB to the primary coolant loop and the pressurizer surge line. Of concern were the material properties for the carbon steel for the primary coolant loop, estimation of the crack opening area at the pipe-to-nozzle interface considering the asymmetry, and the leakage crack size which barely meets the required margin of 2 for the surge line, etc. Some additional work was required by the safety authority to maintain the global safety of the plant at a sufficient level. This paper describes the regulatory application of LBB in Korea, and the issues encountered during the regulatory review.

A Model for Pulsed Activation Accounting for Circulation, Extraction, and Makeup

A model has been developed to accurately calculate the nuclide inventories of the target constituents of Inertial Fusion Energy (IFE) reactors such as HYLIFE-II. It can explicitly account for (1) the combined effects of activation during target implosion (by a high-amplitude flux) and while passing through the reactor chamber (by a low-amplitude flux); (2) decay during circulation in the primary coolant loop, after extraction from the coolant loop, and before re-insertion into the reactor chamber as a new target; (3) continuous extraction and feed-in of target material; and (4) replacement of part of the activation products by makeup materials. The solution strategy uses transition factors – the ratio of the amount of created nuclides to the initial amount – for each system component.

Fuel-failure detection system for Pakistan research reactor-1

Fuel-failure detection in nuclear reactors is important from the safety point of view. After the conversion and upgrading of Pakistan Research Reactor-1 (PARR-1), it was decided to install a fuel-failure detection system to confirm the performance and integrity of the new fuel elements. The fuel-failure detection is based on monitoring of delayed neutrons emitted from fission products leaking into the primary coolant loop from the fuel. For this purpose, two neutron detectors (BF 3) were placed in the graphite moderator blocks that were installed at the outlet coolant pipe in the valve pit. The graphite blocks were enclosed in a lead shield and paraffin wax to reduce background due to gamma radiation and scattering neutrons. A window was provided in the paraffin wax to extract the neutron signal. A gamma spectrometry system including an HPGe detector was also used for determination of the isotopic concentration of fission products. The fuel-failure detection system was tested and calibrated at a miniature neutron source reactor (PARR-2) which provided the basis for alarm limits setting in the event of fuel failure.

Determination of the deuterium content in RBMK-1000 coolant by the method of activation analysis on heavy ions

Activation analysis was performed on reactor coolant samples at different operating powers to determine the deuterium content. The analysis was performed to gain insight into the formation of deposits in, and subsequent corrosion of, primary coolant loops. Methods were developed to perform activation analysis of irradiated samples. The measurements showed that the background was not exceeded in the four samples analyzed for any gamma peak of the radionuclides. The reproducibility and limit of detection of deuterium in water were calculated. The computed excess content of deuterium atoms in the coolant, as compared to the natural content, did not directly depend on the power generated. 7 refs., 2 tabs.

Thermal-Hydraulic Analysis of the Advanced Neutron Source Reactor Refueling Process

The Advanced Neutron Source Reactor (ANSR) is being designed to become the world's most advanced reactor for neutron scattering research. It is both moderated and cooled by heavy water (D[sub 2]O) and uses a core of highly enriched uranium silicide fuel in a plate geometry. During the refueling process of the ANSR, the spent fuel must be moved from within the primary coolant loop (containing D[sub 2]O), through a pool of D[sub 2]O, and finally into a light water spent-fuel storage area. This is accomplished by using a refueling tunnel and fuel transfer cask or lock. Some means of cooling the core, using either natural circulation or forced convection, must be accommodated during this process. Several thermal-hydraulic aspects of this refueling process have been studied. A modified version of the NATCON computer code developed at Argonne National Laboratory was used for the analysis. The NATCON code was revised and modified to incorporate improved friction and heat transfer correlations, and routines for the physical properties of D[sub 2]O were added. The revised code can also accommodate a two-dimensional power density distribution. The results for the refueling process design show that [approximately]48 h is required to ensure the no-boiling condition. The additionmore » of a short chimney reduces that time to [approximately]12 h.« less

Transient Thermal Analysis of a Space Reactor Power System

Space nuclear power systems utilize materials and processes that are completely different from terrestrial reactor systems. Therefore, the tools used to analyze ground-based systems are inappropriate for space reactor design and analysis. The purpose of this study was to develop a space reactor transient analysis tool and to apply this tool to scenarios of interest. The scope of the simulation includes the thermal and neutronic behavior of a liquid-metal-cooled fast reactor, the electrical and thermal performance of the thermoelectric generators, the thermal dynamics of heat pipe radiators, and the thermal behavior of the coolant piping between major components. The thermal model of the system is explicitly coupled to a momentum model of the primary and secondary coolant loops. A one-dimensional conduction model is employed in all solid component models. The reactor model includes an expression for energy generation due to fission and decay heat. The thermoelectric heat exchanger model accounts for thermal energy conversion to useful electrical output. The two-node radiator heat pipe model includes normal operation as well as limited heat pipe operation under sonic limit conditions. The reactor, thermoelectric heat exchanger, and heat pipe models are coupled explicitly by the coolant piping thermal model. The computer program ismore » used to simulate a variety of transients including reactor power changer, degradation of the radiator, and a temporary open circuit condition on the thermoelectrics.« less

Doses from accidental releases of tritium and activation products into the atmosphere

In view of public acceptance and the licensing procedure of projected fusion reactors, the release of tritium and activation products during normal operation as well as after accidents is a significant safety aspect. Calculations have been performed under accidental conditions for unit releases of corrosion products from water coolant loops, of first wall erosion products including different coating materials, and of tritium in its chemical form of tritiated water (HTO). Dose assessments during normal operation have been performed for corrosion products from first wall primary coolant loop and for tritium in both chemical forms (HT/HTO). The two accident consequence assessment (ACA) codes UFOTRI and COSYMA have been applied for the deterministic dose calculations with nearly the same input variables and for several radiological source terms. Furthermore, COSYMA and NORMTRI have been applied for routine release scenarios. The paper analyzes the radioation doses to individuals and the population resulting from the different materials assumed to be released in the environment.

PIUS - status and perspectives

Today's environmental concerns show that nuclear energy is an important option for meeting future increases in global energy demand. Significant nuclear expansion will probably require new reactor designs in which safety is ensured by simple, convincing means. PIUS represents such a reactor design. It is a re-configured 600 MWe PWR, in which the primary safety goal, protection of the reactor core integrity, is entrusted to built-in, self-protective, passive features, without reliance on any monitoring, detection or actuation system, nor operator action. Its basic design features a core that is openly connected, in a natural circulation loop, to a large pool of borated water. The pool is enclosed in a prestressed concrete pressure vessel provided with redundant leakage barriers. The reactor coolant pumps are operated in such a way that there is hydraulic balance in the openings between the primary coolant loop and the pool. Thereby, the hot, low boron content primary loop water is kept separated from the pool water, in spite of the always open natural circulation loop. In severe transients this balance is disturbed, and pool water ingress occurs, shutting down the reactor, or restricting the power to a safe level. The decay heat is transferred to the pool by the natural circulation loop, and a passive pool cooling system, utilizing natural circulation and natural draft cooling towers, prevents boiling of the pool water, even in a station blackout situation. Transient analyses have shown that this passive long-term RHR function will be available in all accident situations, even after double-ended cold leg breaks. Such breaks result in a temporary pressurization of the reactor containment, but the releases of radioactivity will be extremely small and the doses at the fence boundary very low. Cost estimates indicate that PIUS will be quite competitive, and evaluation studies are now under way in several countries.

Conceptual design of a computer-driven display for monitoring reactor coolant mass

The conceptual design of a display for monitoring reactor coolant mass in the primary coolant loop of a pressurized water reactor is discussed. A means-end relational network serves as the framework for decomposing the abstract-the conservation law of mass-into system and process variables. A computer-driven display synthesizes, in real time, mass sources, mass inventory, and mass sinks based on measured plant variables. The purpose of the display is to aid human users in the cognitive tasks of monitoring operations, diagnosing system faults, and responding to component and system failures that affect mass.< <ETX xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">&gt;</ETX>

The PIUS Pressurized Water Reactor: Aspects of Plant Operation and Availability

AbstractThe process inherent ultimate safety (PIUS) reactor is a 600-MW(electric) pressurized water reactor based on a concept developed and verified at ABB Atom during the last 10 years. It is designed to eliminate any possibility of a core degradation accident. Its basic design features a core that is openly connected, in a natural-circulation circuit, to a large pool of heavily borated water. This pool is kept in place by a prestressed concrete pressure vessel provided with redundant leakage barriers. The coolant pumps are operated so that there is hydraulic balance in the openings between the primary coolant loop and the pool. Therefore, the hot, low-boron primary loop water is kept separated from the pool water in spite of the always open natural-circulation path. In severe transients, such as loss of feedwater, this balance is affected, and pool water ingress occurs. Reactor shutdown and long-term residual heat removal are ensured without monitoring and intervention. Thus, safety is independent of p...

Application of leak-before-break to primary loop piping to eliminate pipe whip restraints in a Spanish nuclear power plant

The Spanish plant described in this study is a 982 MWe PWR plant with a three-loop primary circuit of piping made from centrifugally-cast stainless steel SA351 CF8A. The licensee requested an exemption to GDC-4, from CSN, so as to avoid the postulation of guillotine rupture of the primary loop piping. The request was based on the generic work performed for a US PWR plant group to have such exemption to GDC-4. As the piping material in the Spanish plant is different from that in the plants included in the generic work, CSN performed a review of the applicability of the generic results to the Spanish plant. Also, aspects such as fatigue evaluation, net section collapse, crack growth and leak detection, specifically analyzed for the Spanish plant, were reviewed. CSN found that fracture toughness test results from generic work are applicable to the Spanish plant; sufficient margin exists against unstable crack extension, and adequate leak detection capability exists with the leakage detection systems available in the plant. Exemption to GDC-4 was approved and CSN authorized the licensee to remove protection devices against dynamic loads from guillotine breaks in the primary coolant loops.

The safety designs for the TITAN reversed-field pinch reactor study

TITAN is a study to investigate the potential of the reversed-field pinch concept as a compact, high-power density energy system. Two reactor concepts were developed, a self-cooled lithium design with vanadium structure and an aqueous solution loop-in-pool design, both operating at 18 MW/m 2 . The key safety features of the TITAN-I lithium-vanadium blanket design are in material selection, fusion power core configuration selection, lithium piping connections, and passive lithium drain tank system. Based on these safety features and results from accident evaluation, TITAN-I can at least be rated at a level 3 of safety assurance. For the TITAN-II aqueous loop-in-pool design, the key passive feature is the complete submersion of the fusion power core and the corresponding primary coolant loop system into a pool of low temperature water. Based on this key safety design feature, the TITAN-II design can be rated at a level 2 of safety assurance.

Effect of Loop Seal on Reflood Phenomena in PWR

The blockage of the primary coolant loop with water filled at the crossover legs, denoted “loop seal”, is expected to give a significant influence on core cooling during the reflood phase of a PWR-LOCA. However, the effect of the loop seal has been little investigated. Therefore, it was studied experimentally by using Cylindrical Core Test Facility (CCTF). The loop seal was cleared in a short time (40–60 s) after reflood initiation by pushing-off of stagnating water in the crossover legs due to steam accumulation and resultant pressure increase in the upper plenum. Although the core cooling was degraded during the loop seal period, it recovered after the loop seal clearing. The degradation of core cooling during the loop seal period is considered to be caused by the low core-inlet water flow rate and resultantly by the small volumetric fraction of water in the core. The quantitative estimation about the loop seal effect on the clad temperature was made and it was indicated that the maximum clad temperature would not exceed the allowable upper limit (1,473K) specified in the licensing about the reactor safety even with the loop seal at the beginning of the reflood phase.

Effect of loop seal on reflood phenomena in PWR.

The blockage of the primary coolant loop with water filled at the crossover legs, denoted “loop seal”, is expected to give a significant influence on core cooling during the reflood phase of a PWR-LOCA. However, the effect of the loop seal has been little investigated. Therefore, it was studied experimentally by using Cylindrical Core Test Facility (CCTF). The loop seal was cleared in a short time (40–60 s) after reflood initiation by pushing-off of stagnating water in the crossover legs due to steam accumulation and resultant pressure increase in the upper plenum. Although the core cooling was degraded during the loop seal period, it recovered after the loop seal clearing. The degradation of core cooling during the loop seal period is considered to be caused by the low core-inlet water flow rate and resultantly by the small volumetric fraction of water in the core. The quantitative estimation about the loop seal effect on the clad temperature was made and it was indicated that the maximum clad temperatur...

IPFR, Integrated Pool Fusion Reactor Concept

The IPFR (Integrated Pool Fusion Reactor) concept involves placing a fusion reactor into a pool of molten Flibe. The Flibe serves the multiple functions of breeding, cooling, shielding, and moderating. Therefore, the only structural material between the superconducting magnets and the plasma is the first wall. The first wall is a stand-alone structure with no coolant connection and is cooled by Flibe at atmospheric pressure. There is also no need for a primary coolant loop. The design is expected to improve the safety, reliability, and maintainability aspects of the fusion system.

Modeling of Boron Control During Power Transients in a Pressurized Water Reactor

Accurate control instructions in a reactor control aid computer are included in order to realize the boron makeup throughput, which is required to obtain the boron concentration in the primary coolant loop, predicted by a neutronic code. A modeling of the transfer function between the makeup and the primary loop is proposed. The chemical and volumetric control system, the pressurizer, and the primary loop are modeled as instantaneous diffusion cells. The pipes are modeled as time lag lines. The model provides the unstationary boron distributions in the different elements of the setup. A numerical code is developed to calculate the time evolutions of the makeup throughput during power transients.

Uranium Diffusion in H-451 Graphite

The transport or diffusion of uranium (as a standin for plutonium) was investigated under conditions approximating those of the primary coolant loop in a high-temperature gas-cooled reactor. Profiles were obtained for uranium penetration in H-451 graphite at temperatures ranging from 900 to 1400/sup 0/C. Profile data for given temperatures were considered in terms of the following expression: (erf/sup -1/(1-C/C /SUB o/ ))/sup 2/=X/sup 2//4Dt, where C is the concentration of uranium at time t, for distance x, into the pellet; C /SUB o/ is a constant representing the uranium concentration at x = 0 for all t, and D is the diffusion coefficient. Diffusion coefficients for uranium initially present as dicarbide at 1000 and 1400/sup 0/C were found to be defined by Duc/sub 2/ = 3.5 X 10/sup -3/ exp(-4.8 X 10/sup 4//RT) cm/sup 2//s. For uranium initially present as dioxide at 900, 1000, and 1400/sup 0/C, diffusion coefficients are defined by Duo/sub 2/ = 2.34 X 10/sup -6/ exp(-3.2 X 10/sup 4//RT) cm/sup 2//s, where R is the gas constant and T is the temperature in degrees Kelvin.

Critical parameter study of PWR primary coolant pipe leak failure using probabilistic fracture mechanics

This paper discusses the critical parameters which influence the failure probabilities of a PWR primary coolant loop. Probabilistic fracture mechanics (PFM) is applied for the parametric study, using the Monte Carlo program PRAISE to predict the failure probabilities of a PWR primary coolant loop from various distributions of input parameters. Parameters such as nondetection probability of preservice and inservice inspection, vibratory stress, residual stresses, and their correlations are extensively studied. Critical crack depth which causes immediate failure are calculated in the presence of various vibratory stresses with and without residual stresses. Crack growth schemes are determined with various initial defect depth and depth-length ratio as a function of plant operation time. The results show quantitatively how PWR primary coolant loop reliability can be greatly improved by increasing the sensitivity and decreasing the uncertainty of preservice nondestructive inspection.

Analytic Redundancy for On-Line Fault Diagnosis in a Nuclear Reactor

A computer-aided diagnostic technique has been applied to on-line signal validation in an operating nuclear reactor. To avoid installation of additional redundant sensors for the sole purpose of fault isolation, a real-time model of nuclear instrumentation and the thermal-hydraulic process in the primary coolant loop was developed and experimentally validated. The model provides analytically redundant information sufficient for isolation of failed sensors as well as for detection of abnormal plant operation and component malfunctioning. Nomenclature B =bias for sensor calibration b = error bound for measurement C = specific heat F = mass flow rate of primary coolant H = measurement matrix K = product of heat transfer coefficient and area £ = number of measurements M = thermal mass m = measurement p = parity vector Q = power or rate of energy flow S = scale factor for measurement T = temperature t = time V = projection matrix v = sensor output in volts w = weighting coefficient (0 < w < 1) x = true value of a measured variable e = measurement noise 77 = parameter associated with heat transfer £ = shim blade position r = time constant X = fraction of neutron power

Design and implementation of a simple nuclear power plant simulator

A simple PWR nuclear power plant simulator has been designed and implemented on a minicomputer system. The system is intended for students use in understanding the power operation of a nuclear power plant. A PDP-11 minicomputer calculates reactor parameters in real time, uses a graphics terminal to display the results and a keyboard and joystick for control functions. Plant parameters calculated by the model include the core reactivity (based upon control rod positions, soluble boron concentration and reactivity feedback effects), the total core power, the axial core power distribution, the temperature and pressure in the primary and secondary coolant loops, etc.