Developments in Practical Time Series Analysis Methods in the Physical Sciences

A calculation model was developed to predict shutdown dose rate around the recirculation pipes and components in boiling water reactors (BWRs) by simulating the generation, transport, activation and deposition of corrosion products in the primary cooling water. The model is characterized by separating cobalt species in the water into soluble and insoluble materials and then calculating the deposits on the pipe wall for each species using the following considerations:

1. Soluble cobalt (designated as ionic cobalt) is taken into a spinel structure on the surface after diffusing into the oxide layers.

2. Insoluble cobalt (designated as crud cobalt) deposits on the oxide layers.

3. A part of the ionic cobalt released from the crud on the oxide layers is taken into a spinel structure like (1).

The calculated results agree satisfactorily with measurements in two BWR plants. The shutdown dose rate around the recirculation pipes during the entire operating life was calculated to evaluate the effects of the radiation reduction procedures.     

Water Quality Control in Primary Cooling System of Crud Concentration Suppressed Boiling Water Reactor, (I)

The No. 2 Unit of Fukushima-Daini Nuclear Power Plant (2F-2; 1,100 MWe) was commercially operated for 10,320 effective full power hours (EFPH) as its first fuel cycle. The basic design concept of the 2F-2 incorporated the following two features:

(1) Application of procedures for reducing shutdown dose rate based on the Japanese Improvement and Standardization Program

(2) Low crud generation to minimize radioactive waste by careful material selection for the primary system.

Thus, it was possible to keep the average Fe concentration in the condensate water at less than 6 ppb during the first fuel cycle. As a result of this low value, the average life of powdered resin precoated prefilters was extended to about a month, and the average chemical regeneration period of the deep bed demineralizers was extended to more than one year.

The water chemistry of the 2F-2 was characterized by low ⁶⁰Co and high ⁵⁸Co radio-activities in the reactor water, which resulted in a low shutdown dose rate determined mainly by ⁵⁸Co depositing on the primary piping. For example, average dose rate around the primary piping just after reactor shutdown was about 70 mR/h, about 75% of which was from ⁵⁸Co depositing on the pipe inner surfaces. The contribution of ⁶⁰Co was about 25%.     

Effects of Organic Impurities on BWR Water Chemistry, (2)

In order to keep ⁶⁰Co radioactivity low without incurring serious disadvantages regarding integrity of fuel cladding and reliability of structural materials, it is desirable to relate cooling water (especially ⁶⁰Co behavior), structural materials and fuel cladding quantitatively. Organic impurities, such as fragments of cation resin, are fed into the reactor where they decompose, increasing conductivity of the reactor water and decreasing pH, which, in turn, increase ⁶⁰Co radioactivity in the reactor water and then, shutdown dose rate. The relationships between ⁶⁰Co radioactivity in the reactor water and pH have been analyzed by using water chemistry experience at operating BWR plants as well as laboratory data of resin decomposition in the water at elevated temperature, and analytical formulas to explain the shutdown dose rate increase caused by organic impurities intrusion were proposed. Effects of organic impurities on the shutdown dose rate can be moderated preventively by application of weak alkali control.     

Deposition of Nickel and Cobalt Ions with Iron Crud on Heated Stainless Steel Surface under Nucleate Boiling Conditions

Deposition behaviors of Ni (II) and Co (II) ions on a heated surface using simulated Fe crud which was mainly composed of amorphous Fe(III) hydroxide have been studied under nucleate boiling conditions at 553 K and 70 atm. The deposition process of Ni (II) and Co (II) ions is divided into two stages. The first stage is the deposition of hydroxide precipitate on the heated surface by microlayer evaporation and drying out in the nucleate boiling bubble. The second is settlement by conversion of hydroxide into oxides such as NiO, NiFe₂O₄, CoO and CoFe₂O₄. The effective deposition coefficients of Ni(II) and Co (II) ions, without supplying Fe crud, are smaller than that of α-Fe₂O₃ because of the solubility of those hydroxides at a low concentration condition. Their effective deposition coefficients increase with the simulated Fe crud concentration in feedwater and saturate at a value of 0.3 which is calculated theoretically, because the Ni and Co hydroxides react with the simulated Fe crud to produce insoluble NiFe₂O₄ and CoFe₂O₄ on the heated surface. The reaction of Ni deposit with α-Fe₂O₃ does not proceed, but NiO is produced. The reaction rate of Ni deposit with the simulated Fe crud on the heated surface is higher than that of Co deposit.     

Inhibition of Radioactive Contamination of Stainless Steel in Boiling Water Reactor

In order to investigate the later phase of a sodium-water reaction (SWR) event, the code SELPSTA (Sodium-water reaction Event Later Phase System Transient Analyzer) has been developed and the analysis for the long-term system dynamic responses of a SWR event in KALIMER (Korea Advanced Liquid MEtal Reactor) has been made. The SELPSTA code uses the very simple analysis model applied only to the reaction period characterized by a bulk motion, and makes the very quick and concise computation possible. The code reasonably predicts the quasi-steady system transients and has the superiority in the aspect that the various design parameters or operational characteristics are flexibly applicable. In the long-term period of a SWR event, the system dynamic responses analyzed by the code totally depend on the system design parameters such as the breaking pressure of the rupture disk, the variation of the steam injection rate and the sodium drain tank pressure,etc. Based on these analyses results, it is expected that the numerical quantification method of the SELPSTA code is practicable for the long-term system transient analysis and also makes the design of a pressure relief system against a SWR event in a liquid metal reactor (LMR) possible.     

Deposition of Nickel and Cobalt Ions on Heated Surface under Nucleate Boiling Condition

The depositionof Ni and Co ions on a heated surface of simulating fuel rods has been studied in water at atmospheric and 70 atm pressures during nucleate boiling.

The effects of various factors, including heat flux of the heated surface, concentration of coexisting iron oxide (αFe₂O₃ and concentration of Ni and Co ions, on their deposition rate have been investigated. The model for iron oxide deposition which is based on microlayer evaporation and drying out phenomena in the nucleate boiling bubble was shown to be applicable to the deposition of Ni and Co ions. That is, dW/dt=K.Q.C/L, where dW/dt is the deposition rate, K the deposition rate coefficient, Q the heat flux, C the ion concentration, and L the latent heat of vaporization. The K value of Ni ion is about 0.1 and independent of iron oxide concentration. On the hand, the K value of Co ion increases with iron oxide concentration and seems to approach that of iron oxide concentration and seems to approach that of iron oxide (0.3). The Co ion deposited with iron oxide forms Co ferrite. Solubility of Co ferrite is small compared with that of Co deposits without iron oxide (CoO or Co(OH)₂). The increase in the K value of Co ion with iron oxide concentration is attributed to the change in chemical form of Co deposits into more stable species not favoring Co release.     

Hydrazine and Hydrogen Co-injection to Mitigate Stress Corrosion Cracking of Structural Materials in Boiling Water Reactors, (III)

The effects of hydrazine on the corrosion of Zircaloy-2 were examined in supercritical water. Hydrazine could be used as a reducing agent to control the corrosive environment for the coolant of boiling water reactors (BWRs). Before the corrosion test, the applicability of supercritical water for corrosion testing of zirconium alloys was studied. Supercritical water was found to be a useful solvent for testing corrosion based on the following facts: (1) the weight gain of Zircaloy-2 in supercritical water followed the same cubic law with the activation energy of 133 kJ/mol as that in water and steam did, and (2) the weight gain in supercritical water at 723 K and 24.5 MPa was more than 8 times greater than that in water at 561 K and 7.8 MPa depending on immersion time. The corrosion tests in supercritical water at 723 K and 24.5 MPa under γ-irradiation for 1,000 h were conducted to study the effects of adding nitrogen and ammonia on the corrosion of Zircaloy-2. Nitrogen and ammonia are decomposed products of hydrazine. The measured weight gain, oxide film thickness, and amount of hydrogen pick-up had slight differences between cases with and without the additives. Based on these data, it was concluded adding hydrazine to the coolant has little influence on the corrosion of Zircaloy-2 used in BWR cores.     

Some Aspects of LWR's Water Chemistry

The hypothesis laid down in this paper offers an alternative to the current interpretation of the processes: hideout (hideout-return), crud deposition and change of the coolants activity level in the nuclear power reactors under different operating conditions.

This alternative is based on the supposition that the heat flux has not a direct effect on the processes mentioned above, but acts through the heat transfer mechanism in the boundary, caused by itself.

The boundary influenced by heat flow is in non-equilibrium state and in such system states (at adequate heat flux) non-equilibrium structures called dissipative structures arise which ale closely connected with heat transfer mechanism. The transport and the location of the colloidal corrosion products dispersed in steam generators-or reactor water during the units operation are strongly influenced by the existence of dissipative structures. The transport and location of the main part of ion species depend also on the existence of these structures because the colloidal particles act like collectors of the ions dissolved in the water (The ions are inserted in the colloidal particles double layer).

The hideout and hideout-return phenomena are interpreted as closely connected with the existence of the above mentioned dissipative structures. It was attempted to consider the changes in nuclide concentrations in the LWR coolant upon start-up and shutdown as hideout respectively hideout-return processes. The recent shutdown chemistry aspects are discussed also.     

Analysis of Operating Experience Involving Loss of Decay Heat Removal during Reactor Shutdown in Pressurized Water Reactors

In PWRs, loss of decay heat removal (DHR) during reactor shutdown with the reactor coolant system (RCS) partially drained may result in core boiling in a short time. The subsequent RCS pressurization could prevent water flow into the RCS by gravity feed and consequently the core would be uncovered. This paper analyzes U.S. PWR operating experience involving the DHR loss in such reduced inventory conditions.

Between 1976 and 1990, reported were a total of 63 loss of DHR events which occurred during reactor shutdown with the RCS inventory reduced. Review of the event reports indicated that many loss of DHR events in reduced inventory conditions resulted from air entrainment into the DHR pumps due to lowering the reactor water level too far, loss of coolant inventory, increased pump flow and so on.

The coolant heatup rates were evaluated for 12 events with use of the data such as the time elapsed from reactor shutdown actually reported. The calculated results were in reasonably good agreement with the observed ones and showed that core boiling would take place within 1 h even if the DHR loss would occur in the late stage of shutdown (for example, 30 days after the shutdown).     

Application of In-Service Temperature Lowering to Reduce Radioactivity Corrosion Product Deposition on Carbon Steel Piping of BWR Residual Heat Removal System

Assessment of plant data and experiments on deposition of ion species on carbon steel were carried out in order to develop suitable countermeasures to reduce RHR (residual heat removal) piping dose rate. It was thought that radioactivity deposits on the RHR piping were mainly from radioactive ion species in the coolant and they were enhanced by the dehydration reaction of corrosion products on the piping. From an evaluation for temperature dependence of the dehydration reaction, it was proposed to lower the start-up temperature of RHR operation as a way to reduce radioactivity deposition. Feasibility studies of improved RHR operation were conducted and test operations were carried out in the Shika Nuclear Power Station Unit 1. Application of the improved RHR operation resulted in a temperature reduction from 150°C to 110°C, and a radioactivity deposition reduction on the RHR piping to one-fifth ofthat in conventional RHR operation. The improved RHR operation has now been applied to more than fifteen Japanese BWRs and significant suppression effects of radioactivity deposition have been observed.     

Assessment of SAFER03 Code Using ROSA-III Break Area Spectrum Tests on Boiling Water Reactor Loss-of-Coolant Accidents

The SAFER03 computer code has a newly developed evaluation model for the analysis of various boiling water reactor (BWR) loss-of-coolant accidents (LOCAs). Analyses of the ROSA-III break area spectrum tests in a recirculation line were performed using the SAFER03 to assess the predictive capability of the code for a BWR LOCA. The ROSA-III test facility at the Japan Atomic Energy Research Institute (JAERI) was constructed to simulate a LOCA in a BWR/6-251 plant with 848 fuel bundles and 24 jet pumps. This paper summarizes the assessment results of SAFER03 which predicted the system responses and key phenomena well and the conservative peak cladding temperature (PCT) for recirculation line break tests with different break areas.     

Cobalt Deposition and Release on Magnetite Powder in High Temperature Water

By making use of miniature ²³⁹Pu and ²³⁵U fission chambers and an accelerator-based neutron source, the ²³⁹Pu/²³⁵U neutron fission rate ratio profile has been measured in the region of a plane interface between a tank containing ordinary water and a tank containing an aqueous solution of boric acid. The reported preliminary results indicate an approximately 8 mm shift of the central point of the fission rate ratio profile towards the poisoned side of the system and an accompanying larger shift in the ²³⁵U Cd ratio profile towards the unpoisoned side.     

Effects of ECC Water Injection Rate on Reflood Phase during LOCA

Some metal iodides such as of Fe, Al, Zr and Te are known to cause stress corrosion cracking (SCC) of Zircaloy just as iodine itself does. Therefore 15 metal iodides were selected as corrodants, and SCC tests were carried out using the internal gas pressurization method.

The results showed that: (1) only those metal iodides which react thermodynamically with Zr to produce ZrI₄ cause SCC of Zircaloy-2; (2) when SCC occurs, the reaction rate between the iodide and Zr seems to be a main factor in determining the SCC susceptibility; (3) gaseous ZrI₄ is the most corrosive agent; and (4) some species of metal iodides, such as PbI, cause SCC of Zircaloy-2 more easily than I₂ vapor.

Scanning electron microscope (SEM) examination and electron probe microanalysis (EPMA) on the fracture surface of failed specimens revealed that ZrI₄, formed as the reaction product between the metal iodides and Zr, might induce SCC of Zircaloy-2 rather than the iodides themselves.     

Temperature Dependence of Cobalt Release Rate from Stainless Steel in Neutral Water

In order to evaluate the release source of Co from heated tubes of the feedwater line of BWR, Co release rate measurements were carried out by measuring ⁶⁰Co released from irradiated stainless steel in contact with neutral water at an oxygen concentration of 20 ppb. The temperature- and time-dependences on Co release rate were studied for 670 h release experiments; the rate was found to decrease in proportion to t ^?frac12 (t ^?frac12 (t: exposure time) and to have its maximum value around 240°C. An empirical equation was obtained for the Co release rates in the temperature range of 150~240°C.

R_Co = 1.9 × 10⁶ exp (-14,300/(RT)) t ^?frac12

where R _co Co release rate (mg/(dm²·month)), R: Gas constant (cal/(mol·deg.)), T: Exposure temperature (K) and t: Exposure time (h). Decrease in the Co release rate with temperatures over 250°C, seemed to be ascribed to the formation of a protective oxide layer on the surface of stainless steel.     

Optimization of Dissolved Hydrogen Concentration for Control of Primary Coolant Radiolysis in Pressurized Water Reactors

Optimization of the dissolved hydrogen concentration in primary coolant of pressurized water reactors has several potential advantages in material integrity and dose reduction. To assess the threshold value of dissolved hydrogen for reducing condition, in-pile loop experiments and radiolysis model calculations were performed. Both experiments and model calculations indicate that the threshold value of dissolved hydrogen for radiolysis is much less than the present control level in pressurized water reactors, and the in-core region is more easily affected by reduced dissolved hydrogen level than the out-of-core region. The measurement of electrochemical corrosion potential of structural material, while varying dissolved hydrogen levels, is desirable in the future to compensate for the uncertainty of the model calculation. For this purpose, electrochemical corrosion potential of stainless steel at the top-of-core, in which the reference electrode could be inserted, was estimated based on the correlation between the experiments and the modeling for the loop.     

Effects of Hydrogen Peroxide on Intergranular Stress Corrosion Cracking of Stainless Steel in High Temperature Water, (V)

The difference in electrochemical corrosion potential of stainless steel exposed to high temperature pure water containing hydrogen peroxide (H₂O₂) and oxygen (O₂)is caused by differences in chemical form of oxide films. In order to identify differences in oxide film structures on stainless steel after exposure to H₂O₂ and O₂ environments, characteristics of the oxide films have been examined by multilateral surface analyses, e.g., X-ray diffraction (XRD), Rutherford back scattering spectroscopy (RBS), secondary ion mass spectroscopy (SIMS) and X-ray photoelectron spectroscopy (XPS). Preliminary characterization results of oxide films confirmed that the oxide film formed under the H2O2 environment consists mainly of hematite (α-Fe₂O₂), while that under the O₂ environment consists of magnetite (Fe₃O₄). Furthermore oxidation at the very surface of the film is much more enhanced under the H₂O₂ environment than that under the O₂ environment. It was speculated that metal hydroxide plays an important role in oxidation of stainless steel in the presence of H₂O₂. The difference in electric resistance of oxide film causes the difference in anodic polarization properties. It is recommended that several anodic polarization curves for specimens with differently oxidized films should be prepared to calculate ECP based on the Evans diagram.     

Chemical State and Deposition Rate of Nickel Ion Deposit Produced on Heated Surface in High Pressure Boiling Water

Nickel ion deposit was produced on a heated rod surface in high pressure boiling water (150–285°C, 0.4–7.0 MPa). The deposit under the same temperature and pressure conditions as those for BWR reactor water (285°C, 7.0 MPa) was identified as NiO by spectrum profile analysis of the NiLα, NiLβ and 9th-order NiKα₁ lines. Deposition rate was obtained from in situ measurements of deposit thickness, by a photoacoustic method, and from chemical analysis of deposit amount. The deposition rate coefficients obtained in temperature and pressure ranges of 150–250°C and 0.4–4.0 MPa were 2 × 10^?3–5 × 10^?2, which were 0.15–0.45 times as large as those of iron crud. This was attributed to a dissolution effect of Ni ion from NiO. The deposition rate coefficient at 285°C, 7.0 MPa was estimated to be 4.4 × 10^?2–1.3 × 10^?1.     

Experimental Analysis on Mitigation of Thermal Stratification in the Suppression Pool of a Water Wall Type Passive Containment Cooling System

A water wall type passive containment cooling system with an outer pool surrounding the suppression pool is one passive containment cooling system. In the system, a baffle plate in the suppression pool mitigates thermal stratification formed at the vent tube outlet level and enlarges the heat transfer area. The effectiveness of the baffle plate in mitigating thermal stratification was experimentally confirmed; the heat transferred to the outer pool increased about 50% due to a larger high temperature region and a longer effective heat transfer length. The experimental analysis was performed using a three-dimensional thermal-hydraulic analysis program. In the analysis, a laminar flow model and slip conditions on structural walls were used, and the calculated temperature profiles and natural circulation flow rates along the baffle plate agreed with measurements. The model was then judged as a valid and practical tool to evaluate global natural circulation and temperature distributions in a large pool. And it was analytically corn- firmed that the thermal resistance of the PCV wall and the heat flux to the outer pool affected the performance of the baffle plate.