Transport and Deposition of Metals in Sodium-Stainless Steel Systems, (II)

Making use of the radioisotopes of manganese, zinc and antimony, to represent radioactive corrosion products, and a radioisotope of cesium as representative fission product, the transport and deposition of these radioisotopes in a natural circulation sodium loop have been studied over a temperature range from 630°C down to 280°C.

The movement of the radioisotopes during the loop running were monitored byin situ counting and radiochemical analysis of sodium samples taken from the loop. The distribution of the radioisotopes over the various cross sections along the cooling tube were determined after quick solidification of flowing sodium by chilling.

The results indicate that manganese deposits on the substrate materials presumably by adsorption in the higher-temperature region and by precipitation in the lower-temperature region, zinc presumably by adsorption associated with plating in the lower-temperature region, antimony by coprecipitation with other deposits in the cold settling pocket, and cesium mainly by physical adsorption in the lower- temperature region.

Significant deposition of the radioisotopes was observed in the pocket and the apparent rate constants of the deposition were obtained for manganese, zinc and cesium.     

Transport and Deposition of Metals in Sodium-Stainless Steel Systems, (IV)

With the object of gaining information on the behavior of radioactive corrosion products such as ⁵⁴Mn, ⁶⁵Zn and ¹²⁴Sb in relation with oxygen concentration in liquid sodium-stainless steel systems, deposition experiments were undertaken with use made of stainless steel capsules, each containing 20 g of sodium and a trace amount of one of these radioisotopes.

Each capsule was set vertically and subjected to a steep temperature gradient, descending toward the bottom, for 700 hr of run to let the sodium oxide diffuse through the liquid sodium and settle on the bottom. The oxygen concentration in the sodium was controlled by varying the bottom temperature. The concentrations of the radioisotope at different radial positions on various cross sections along the capsule axis were determined on sodium samples dissected from the capsule after quick solidification of the sodium by chilling.

The values of the activation energy of deposition were determined for each radioisotope from partition coefficients, defined as the ratio of specific radioactivity between that of the stainless steel surfaces and that in the liquid sodium. Manganese deposition involves precipitation in the lower temperature region, and adsorption—associated loosely with oxygen in the sodium—in the higher temperatures. Zinc adsorption was observed on the wall surface only when it had been previously treated with sodium before dissolution of the radioisotope in the sodium. Antimony deposition is attributable to coprecipitation with sodium oxide.     

Transport and Deposition of Halide in Alkali Metal-Stainless Steel Systems, (II)

Solubility of sodium iodide in liquid sodium was determined over a temperature range of 350–800°C by using equilibration and sampling techniques.

The dissolution of a 2–3 g fragment of sodium iodide crystal into 7.5–8g sodium was made in a base cell of reversed: “J” type-stainless steel capsule, which was heated in an electric resistance furnace. After the equilibration had been established, a portion of the solution was decanted into a side arm cell of the capsule by tilting the capsule about 110° with the furnace capable of rotating round its horizontal axis. The sodium transferred to the arm cell was analyzed as a single sample.

The solubility: S in the unit of ppm by weight is expressed as a function of temperature in the unit of Kelvin in the forms

• log₁₀S=8.46—3,440/T above 660°C

• log₁₀S=9.65~4,550/T below 660°C

which fit well with Bredig's data in the higher temperature region. The partial molar excess quantities, calculated from the thermodynamic properties of sodium iodide and its individual solubility data, are in reasonable agreement with those determined for the dilute solution of the iodide in liquid sodium by Castleman & Tang.     

Cold Trapping of Fission Products in a Stainless Steel Sodium Loop

The behavior of fission products in sodium system was studied using a cold trap installed in the Toshiba Fission Product Loop, which is a stainless steel inpile sodium loop. The fission products were generated within the liquid sodium by irradiating UO² fuel under sodium convection. The cold trapping behavior of ⁹⁵Zr, ⁹⁵Nb, ¹³²Te, ¹³¹I ¹³⁷Cs and ¹⁴⁰La (¹⁴⁰Ba) was examined by measuring the γ-rays in the cold trap and in the sodium dump tank by means of Ge(Li) detectors. Such fission product nuclides as ^{131 132}Te and ¹³⁷Cs were found to be reversively removed by the cold trap at lower temperatures, with negligible hysteresis. The distribution coefficient K(cm) for the cold trap surface was found tobe expressible by log K=(3.08±0.17)-(0.0112±0.0006)×T(°C) for ¹³¹I, and by log K:=-(4.55±0.29) + (1830±1/140) × 1/T(K) for ¹³⁷Cs. The behavior of ¹³¹I and ¹³⁷Cs during cold trapping can be explained clearly by adsorption model using these K values. While ⁹⁵Zr, ⁹⁵Nb, ¹⁰³Ru and ¹⁴⁰La(¹⁴⁰;Ba) also were detected in the cold trap, the efficacy of cold trapping is far lower for these nuclides than for ¹³¹I and ¹³⁷Cs, and changing the cold trap temperature did not appear to have any appreciable effect on the cold trapping. Presence of ¹⁴¹Ce and ¹⁴⁴Ce was not detectable in the cold trap, although small amounts of these nuclides were detected in the sodium dump tank.     

Transport and Deposition of Halide in Alkali Metal-Stainless Steel Systems, (III)

Solubility of sodium iodide in liquid sodium was determined over a temperature range of 350~650°C by using capsules of two different kinds.

Stainless steel capsule of short type loaded with 2~3.5g sodium and fragments of sodium iodide crystal was put in a pair of holders and heated under an uniform temperature distribution in a vertical electric furnace. Stainless steel capsule of long type loaded with 20 g sodium and powders of the iodide was heated at its upper part in the furnace and cooled at its bottom to establish a large temperature gradient along the capsule tube. After a given term of heating, the iodide was fixed in solidified sodium by quick quenching of these capsules.

The iodide solubility data measured by using the capsules are in good agreement with the data reported by Miyahara, Power Reactor and Nuclear Fuel Development Corp. (PNC). It is not observed in the measurement that the iodide is enriched at free surface of the liquid sodium and forms particles in the solution; these were thought of as reasons why PNC expressions of the solubility are about 10 times as large as Allan's expression.     

Diffusion of Cesium in Stainless Steel and Possible Implications for Chromium Depletion and Mobility

Corrosion of the steel cladding of mixed oxide fast breeder fuel elements is thought to be initiated by penetration of Cs (and oxygen) into the grain boundaries of the steel. Experiments were therefore performed to investigate the penetration of Cs into stainless steel (1.4970 and 1.4988) with the aid of radioactive tracers and with Rutherford backscattering techniques. The penetration distances were very limited and decreased with decreasing oxygen potential. Chromium diffusion was also measured in one of the steels (1.4970) and yielded D ^Cr-values of the order of magnitude of reported D ^Cs-values. Chromium diffusion along grain boundaries was enhanced in steel preloaded with Cs.     

Some Experiments and a Practical Method on Dosimeter Glass Washing Process

With the object of gaining an insight into cesium behavior in its relation to oxygen concentration in sodium-stainless steel systems, a deposition experiment was carried out with use made of stainless steel capsules containing 20 g of sodium and a trace amount of radioactive cesium.

The capsule was vertically set to establish a steep axial temperature gradient, descending forward the bottom, and maintained during a period considered ample for the sodium oxide contained in excess of saturation to settle on the capsule bottom. The oxygen concentration in the sodium was varied by adjusting the temperature of the capsule bottom. The radial distribution of cesium across samples sectioned from different parts along the length of the capsule tube were determined on annular samples extruded from sections of the capsule after chilling for rapid solidification of the sodium.

The activation energy for cesium deposition was obtained from partition coefficients—defined as the ratio of cesium concentration on the stainless steel surface and that in the bulk sodium. The resulting data on activation energy indicate that cesium radioisotope deposits in the temperature region below 440°C presumably by physical adsorption on the stainless steel surface when it has previously been exposed to high temperature sodium, and that this surface adsorption appears to be insensitive to differences in oxygen concentration in the sodium.     

Corrosion of Austenitic Stainless Steel in Steam Containing Cesium Hydroxide

Corrosion behavior of SUS-304L SS, which is used as a typical material for reactor structures has been studied under the steam containing CsOH at temperatures of 450~750°C. The steam was supplied by heating the aqueous solution of CsOH with concentrations 0~320 mM at 110°C. The mass gain by oxygen uptake increases parabolically with reaction time. Since the activation energy of reaction rate constant is consistent with that of self diffusion of Fe in Fe₃O₄, it can be considered that the corrosion reaction is determined by the diffusion rate of Fe atoms in the outer oxide layer. The effect of CsOH on the corrosion reaction under the steam is large in the initial stage of the reaction, but small in the final stage. The mechanism of corrosion reaction has been discussed on the basis of microstructure and composition of the reaction layers.     

Investigation of Cobalt Deposition Behavior with Zinc Injection on Stainless Steel under BWR Conditions

Radioactive⁵⁸Co and ⁶⁵Zn were used together to investigate the deposition behavior of cobalt and zinc in the oxide film formed on the stainless steel piping of an apparatus simulating the high temperature and pressure conditions of a boiling water reactor. Using these radioactive tracers allowed the experimental setup to closely approximate the conditions found in actual plants. The accumulation of ⁵⁸Co and ⁶⁵Zn on the stainless steel piping was monitored using an online gamma-ray detector. The results were as follows. At higher zinc concentrations, the early period of fast deposition was shortened and the later deposition phase was slowed. Two mechanisms appeared to be responsible for the suppression of cobalt deposition by zinc injection. The first worked by decreasing the growth rate of the oxide film. This effect appeared most conspicuously in the outer oxide layer under normal water chemistry conditions. The second mechanism worked by reducing the cobalt concentration in the oxide film. This second effect was prominent in the inner oxide layer under hydrogen water chemistry conditions.     

Diffusion of Cesium in Stainless Steel under Condition of Oxygen Potential Controlled by Mo/MoO2 or NbO2/Nb2O5 Oxygen Buffer

The kinetic behaviors of cesium migration in SUS-316 stainless steel for a cladding material of fast breeder reactor were studied using radioactive ¹³⁴Cs as a tracer. The oxygen potential in the atmosphere surrounding the specimen was controlled by Mo/MoO₂or NbO₂/Nb₂O₅ oxygen buffer, which corresponds approximately to the one in fuel-cladding gap during irradiation. The concentration profile of Cs in the specimen was analyzed on the basis of diffusion theory. The temperature dependences of diffusion coefficients of Cs in SUS-316 stainless steel were expressed in the range of 650~800°C by the following equations:

In the oxygen potential controlled by Mo/MoO₂ oxygen buffer

D = 0.15 exp (?63,500/RT) (cm²/s).

In the oxygen potential controlled by NbO₂/Nb₂O₅ oxygen buffer

D = 9.0×10-⁵exp(?50,300/RT) (cm²/s)

where the activation energy is expressed in terms of cal/mol.     

Studies on Sodium Boiling Accidents in a Large Fast Breeder Reactor, (II) Radial Propagation of Sodium Boiling by Reactivity Effect

A computer code BORE was developed, with which analyses were performed on channel plugging accidents that would occur on a 1,500 MWe LMFBR. The BORE code calculated the dynamic characteristics of coolant boiling and fuel failure propagation radially in the core, and the requirements of core instrumentation systems were also analyzed. The results show that coolant boiling and/or fuel failure in a channel plugging accident are propagated only to a limited number of adjacent channels when sensors are installed that detect anomalies in channel flow, channel outlet temperature, boiling or reactivity. It is also concluded that the coolant void effect is not serious from the standpoint of safety when the time required for boiling propagation to adjacent channels can be made longer than 0.15 sec.     

Development of a Suppression Method for Deposition of Radioactive Cobalt after Chemical Decontamination: (I) Effect of the Ferrite Film Coating on Suppression of Cobalt Deposition

In the last decade, chemical decontamination at the beginning of periodical inspection has been applied to many Japanese BWR plants in order to reduce radiation exposure. However, following the chemical decontamination, a rapid dose rate increase can be seen in some plants after just a few operation cycles. Oxide film, which easily incorporates radioactivity, might be formed after the chemical decontamination. We developed a new way to reduce the recontamination after the chemical decontamination to maintain long-term continued decontamination effects without any chemical injections or chemical controls in reactor water during operation. In our approach, a fine ferrite film is formed by the Hitachi Ferrite Coat process after oxide films formed during the plant operation are removed by the chemical decontamination process.We select Fe(HCOO)₂ aqueous solution, H₂O₂, and N₂H₄ as the treatment chemicals for fine ferrite film formation for suitable BWR plant application. Our laboratory experiment results confirm a ⁶⁰Co deposition reduction effect of 1/5 compared with that of nontreatment for up to 3,100 hours. The fine ferritefilm that was formed on the specimen before the ⁶⁰Co deposition test remains as a film structure after the test. The corrosion amount of the specimen is suppressed to 1/4 through the effect of the fine ferrite film.     

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

The stress corrosion cracking (SCC) of structural materials used in boiling water reactors has been studied at relatively low hydrogen peroxide (H₂O₂) concentrations, around lOppb, which was assumed to be representative of the corrosion environment formed in hydrogen water chemistry (HWC). The 1/4T compact tension specimen was used for measurement of crack growth rates (CGRs) of sensitized type 304 stainless steel in high temperature and high purity water. Crack length was monitored by a reversing direct current potential drop method. Since H2O2 is easily decomposed thermally, a polytetrafluoroethylene-lined autoclave was used to minimize its decomposition on the autoclave surface. The CGR in the H₂O₂ environment differed from that in the O₂ environment even though the electrochemical corrosion potential (ECP) for both conditions was the same. The data implied that the ECP could not be used as a common environmental deterministic parameter for SCC behavior at higher potentials for different oxidant conditions. The corrosion current density was found to play an important role as an environmental index for SCC, which was given as just the current density at the ECP at a specific oxidant concentration. The CGRs were found to be written as CGR = (3.8±0.6)xl0^-3 _icor +(l-5±1.6) x 10^-8mm/s using the calculated corrosion current density i_corbelow 10^-4 A-cm^-2.     

Physics Studies on BWR Void Simulated Cores, (I)

In an attempt to obtain more information concerning steam void effects on the physical behavior of BWR cores, voids ranging from zero to 24% volume fractions were statically simulated by inserting polyethylene tubes between fuel rods of simulated BWR core, and measurements were made of local flux distribution, neutron spectrum, reactivity, reactivity coefficient of void and so forth. The results were compared with calculations. Improved methods of calculation were developed, which assured better agreement with measured results.

While this paper only deals with local distributions of the thermal neutron flux and spectrum within the assembly cells, it has been shown that spectrum variation near water gaps could be well described by an overlapping model which provided an improvement on the analysis. An additional improvement was obtained by the incorporation of an approximate transport correction.     

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.     

Compatibility of Eu2O3 and (Eu,Gd)2O3 with Type 316 Stainless Steel

The compatibility of Eu₃O₃ and (Eu,Gd) ₂O₃ mixed oxides which contain 0, 1, 5, 10, 50 and 100 mol% of Gd₂O₃, with type 316 stainless steel was examined by out-of-pile heatings. The heating temperature ranged from 550 to 1,000°C and heating time from 500 h to up to 5,000 h. Metallographic examinations of the stainless steel indicated that chemical reaction between Eu₂O₃ and stainless steel occurred at 650°C and that grain boundary penetration extended up to 40 μ m after heating at 1,000°C for 500 h. The penetration depth Δ was expressed as a function of heating temperature T as

Δ = 3.08 × 10⁴ exp(?1.72 × 10²RT)

where R is the gas constant and the activation energy is given in cal/mol. Electron probe microanalysis and X-ray diffraction analysis indicated that main reaction product is europium silicate. When Gd, which is a decay product of Eu, is mixed with Eu₂O₃, the degree of reaction between (Eu,Gd) ₂O₃ and stainless steel decreases as increase in concentration of Gd₂O₃. It was found that actually no grain boundary attack was observed even after heating at 1,000°C for 500 h when concentration of Gd₂O₃;i in (Eu,Gd) ₂O₃ exceeded 10 Discussion was made of reaction mechanism of Eu₂O₃ and (Eu,Gd) ₂O₃ with stainless steel.     

Diffusion Behavior of Cesium,Strontium and Cobalt in Pellet Waste Package

A leaching model for a radioactive pellet package for nuclear power plant wastes was studied. Calculated results obtained in term of the model were compared with experimental ones, where volumetric averages of diffusion coefficients in the Na₂SO₄ pellet and in the cement-glass were used to solve the diffusion equation in the pellet package. Since Na₂SO₄ pellets dissolved quickly, the pellet part were simulated by saturated Na₂SO₄ solution. The leaching ratios were meaured with different volume ratios of cement-glass in the package to discuss the validity of the model. The cement glass diffusion coefficient and distribution factor, which were required to solve the equation, were also measured. The calculated results agreed with the experimental ones within a 20% error.     

Ti35钛合金与低碳不锈钢在硝酸溶液中的电化学腐蚀对比研究

从电化学角度对Ti35合金(Ti-5%Ta)和超低碳不锈钢在硝酸溶液中的抗腐蚀性能进行了对比评价。对不同温度、不同浓度、不同添加离子等条件下的阳极极化曲线进行了测试。得到结论:在相同温度下,硝酸溶液浓度的变化导致了钛合金的腐蚀加剧。在相同浓度下,随着温度的升高,钛合金腐蚀越来越不明显。Cr6 离子浓度对钛合金腐蚀能力的影响不大。铀酰离子的作用较弱,加了铬和钌离子后表现出对阳极反应的明显抑制作用。通过表面微观形貌的观察,发现不锈钢对硝酸浓度的变化更为敏感,氧化性离子的综合作用对于不锈钢的耐蚀性影响很大。研究结果表明,Ti35合金具有比不锈钢优异得多的抗腐蚀性能,有希望取代000Cr25Ni20不锈钢而用于后处理设备。