JP2009008407A - Fuel assembly - Google Patents

Abstract

<P>PROBLEM TO BE SOLVED: To improve the scram-transient characteristics and stability of a boiling water reactor without lowering the economical efficiency of it. <P>SOLUTION: The percentage of the volume of a deficient region to that in the case where a nuclear fuel material is accommodated all over an active fuel section in all tetragonal lattice positions is set between 3% and 6% inclusive by keeping a fuel assembly where fuel rods are laid out at least a part of tetragonal lattice positions of N rows and N columns equipped with partially deficient fuel rods with deficient regions accommodating no nuclear fuel material which are formed in the range extending downward from the upper end of the active fuel section beyond 1/6 of the active fuel length A. This makes it possible to load more nuclear fuel materials than the loaded volume of a nuclear fuel material of 9 rows and 9 columns (A type) and to improve the scram-transient characteristics such as ΔMCPR and stability. Similar effects can also be obtained by forming a deficient region ranging downward from the upper end of the active fuel section beyond 1/6 of the active fuel length in fuel rods of the number between 1/5 and 1/3 inclusive of N<SP>2</SP>and by other methods. <P>COPYRIGHT: (C)2009,JPO&INPIT

Description

  The present invention relates to a fuel assembly loaded in a boiling water reactor.

  FIG. 52 is a block diagram with a schematic cross section of a portion of a typical boiling water nuclear power plant (BWR). In the BWR, a reactor pressure vessel 2 is provided in a reactor containment vessel 1. The reactor pressure vessel 2 accommodates a reactor core 3 composed of cooling water 4, a plurality of fuel assemblies, control rods, and the like.

  The cooling water 4 is forcibly circulated by the recirculation system 8 and receives heat generated by nuclear fission such as uranium 235 (U235) in the core 3, so that saturated water and saturated steam are mixed, and the core 3 Move to the top. The saturated steam is dried by the steam separator and the steam dryer, and is sent to the turbine 10 via the main steam piping system 9 connected to the reactor pressure vessel 2 to drive the turbine 10. The generator 15 is rotated by driving the turbine 10 to generate electricity.

  The steam that has worked in the turbine 10 is introduced into the condenser 16 to become condensate. After the pressure is increased by the condensate pump 17 and the temperature is raised by the feed water heater 18, the steam is again fed into the reactor pressure vessel 2 by the feed water pump 19. To be supplied.

  FIG. 53 is a diagram showing typical operating characteristics of the BWR. Normal operation is performed on the rated output curve, design flow control curve, stability limit curve, minimum pump speed curve, cavitation limit curve, maximum pump speed curve, in the area surrounded by them, and on the natural circulation curve. Done. In the example shown in FIG. 53, the rated output is achieved when the core flow rate is 85% (point A) to 105% (point B). Reactor operation is usually near the 85% flow rate at the beginning of the cycle, moves to a higher flow rate as the cycle burnup progresses to take advantage of the reactivity gain due to the increased coolant flow rate, and 105% flow rate at the end of the cycle. It becomes a neighborhood.

  In addition, the core nuclear power design is usually performed so that the core axial power distribution has a lower strain at the beginning of the cycle, and the void ratio at the core outlet is increased. This promotes the generation of plutonium 239 (Pu239), which is a fissile material, from uranium 238 (U238). From the middle to the end of the cycle, the operation is to burn Pu239 in addition to U235, so the power distribution in the core axis direction tends to be distorted. In the core with improved economic efficiency, the fuel assembly core design (axial enrichment distribution, number of fuel rods containing Gd, etc.) and the core design (assembly) Body loading pattern, control rod pattern, etc.)

  FIG. 54 is a cross-sectional view schematically showing a typical BWR core. A plurality of fuel assemblies 30 covered with a channel box are loaded in the core 3 of the BWR so as to surround the control rod 32. The channel box is a rectangular tube having a vertically long thin cross section having a substantially square shape. Further, a plurality of local output region monitors (LPRM) 33 are arranged in the core for detecting the neutron flux.

  FIG. 55 is a longitudinal sectional view schematically showing a typical BWR core. Each fuel assembly 30 is supported by a core support plate 34 and an upper lattice plate 35, and is surrounded by a cylindrical shroud 36. The cooling water 4 flows into the channel box 31 from below through the orifice of the fuel support bracket 37 and the lower tie plate, is heated by the fuel assembly 30, generates steam (void) by boiling, It becomes a phase flow. A typical commercial BWR fuel assembly effective length is about 3.7 m.

  As fuel assemblies in such boiling water reactors, 7 rows and 7 columns, 8 rows and 8 columns, improved 8 rows and 8 columns, high burn-up 8 since commercial power generation was performed in Japan. A row 8 column type and a high burnup 9 row 9 column type have been adopted.

  FIG. 56 is a partially cutaway perspective view of a 9 × 9 fuel assembly (type A). FIG. 57 is a partially enlarged perspective view of the 9 × 9 fuel assembly (A type). FIG. 58 is a cross-sectional view showing a 9 × 9 fuel assembly (type A) together with control rods.

  The 9 × 9 fuel assembly (type A) includes a plurality of fuel rods 71 and 81 and a water rod 53 in which a nuclear fuel material pellet 54 is housed in a cladding tube 55. These fuel rods 71 and 81 and the water rod 53 are held at both ends by an upper tie plate 61 and a lower tie plate 62. Further, seven spacers 63 are arranged in the axial direction of the fuel assembly, and the distance between the fuel rods 71 and 81 and the water rod 53 is maintained. In the 9 × 9 fuel assembly (type A), eight partial-length fuel rods 81 whose height does not reach the upper tie plate 61 are employed.

  The fuel assembly 30 is generally loaded in a lattice shape on the core 3 of the nuclear reactor. When the fuel assembly 30 is loaded in a lattice shape, the distribution of water as a moderator is relatively large in the portion (water gap portion) between the channel box 31 adjacent to the water rod 53 portion in the upper part of the core. Become.

  FIG. 59 is a graph showing an example of core average axial power distribution at the beginning and end of a cycle for a core loaded with a 9 × 9 fuel assembly (type A).

  The 9 × 9 fuel assembly (type A) has an increased amount of fissile material per fuel assembly 30 compared to the high burnup 8-row, 8-column type fuel assembly, and the inside of the fuel assembly 30 Appropriate distribution of enrichment and proper arrangement of flammable poisons. Thereby, high burnup and long-term operation cycle are realized, and the economic efficiency of the core 3 is improved.

  The 9 × 9 fuel assembly (type A) has an increased total heat transfer area due to the increase in the number of fuel rods compared to the high burnup 8-row 8-column type fuel assembly, so that the limit output characteristics are improved. . Moreover, the deterioration of stability based on nuclear factors due to the increase in negative void coefficient (or absolute value of void reactivity coefficient) due to high burnup and long-term operation cycle is due to the adoption of two large diameter water rods 53, etc. It is suppressed by.

  The deterioration of the stability based on the thermohydraulic factor due to the increase of the two-phase pressure loss accompanying the increase of the assembly lattice is caused by the reduction of the spacer pressure loss coefficient, the eight partial length fuel rods 81, and the high pressure loss type lower tie plate 62. It is suppressed by adoption of. That is, the reduction of the spacer pressure loss coefficient and the eight partial length fuel rods 81 having the length of about 2/3 of the full length fuel rod 71 (effective length is about 2.2 m) contribute to the reduction of the two-phase flow pressure loss. The high pressure loss type lower tie plate 62 contributes to an increase in single phase flow pressure loss. As a result, the ratio of single-phase flow pressure loss to total pressure loss is increased, and the stability of the core is improved.

  The limit output characteristic means the level of an output threshold at which transition boiling occurs in the fuel assembly under given operating conditions (output, flow rate, pressure, inlet enthalpy). In addition to the number of fuel assembly lattices, the limit output characteristics are also affected by the core fuel lattice shape, the spacer type, the number, shape, and arrangement of water rods.

  For this reason, when developing a new fuel assembly, a thermal hydraulic test is normally performed on the actual size simulated fuel assembly under the BWR operating condition to confirm the limit output characteristics. Furthermore, the test results are organized into a boiling transition correlation type of “limit quality versus boiling length” type by using a thermal hydraulic analysis code for the BWR fuel assembly. This is because in the boiling water reactor, as the boiling length increases, the liquid film thickness covering the cladding tube 55 becomes thinner, and the height at which the liquid film disappears is considered as the transition boiling occurrence point.

  A typical example is the GEXL correlation formula. In the commercial BWR currently in operation, based on the GEXL correlation formula obtained for each fuel assembly type, the minimum critical output ratio (Minimum Critical Power) Ratio: MCPR) and MCPR change amount (ΔMCPR) during transition are often evaluated. The GEXL correlation formula reflects that the margin for the limit power is large in the lower region of the core 3 of the BWR.

  When a turbine trip occurs when the BWR is operating at rated power, the turbine stop valve is closed and the turbine steam bypass valve is opened rapidly, and at the same time, the control rod is rapidly inserted (reactor scram). At this time, when the bypass valve is not operating, the positive surface reactivity is temporarily injected due to the decrease in the void fraction in the core accompanying the increase in the reactor pressure, so the fuel surface heat flux rises transiently. MCPR decreases. The amount of decrease in MCPR is maximized at the end of the cycle when the negative void coefficient increases and the control rod is almost fully pulled out. The BWR transient analysis model is described in Non-Patent Document 1, for example.

  FIG. 60 to FIG. 62 show the results of turbine trip / bypass valve malfunction analysis for a conventional scram BWR plant loaded with all 9 × 9 fuel A types. This analysis result is a result of analysis using the TRACG code.

  Immediately after the turbine stop valve is closed, the neutron flux rapidly rises due to the positive reactivity injection due to the void fraction reduction, but due to the Doppler effect accompanying the fuel temperature rise (negative reactivity injection) and the large negative reactivity injection due to the scram, Bundles will drop sharply. The core flow rate is decreasing because the two recirculation pumps are tripped by the turbine trip signal in order to mitigate the void rate decrease due to the pressure increase. The rise in pressure is suppressed by opening the relief safety valve. Due to the transient increase in the fuel surface heat flux, MCPR temporarily decreases but recovers without decreasing to the safety limit MCPR (1.07) (ΔMCPR = 0.25), and the event does not reach the boiling transition state. Is over.

  Stability refers to a low flow rate / high output state when the plant is started or stopped, or a transient change such as a trip of one recirculation pump occurs in the plant, resulting in a low flow point. / This means the damping characteristics of neutron flux oscillation when shifting to high power. The core 3 is desirably stable in the entire operation region, and is confirmed by indicating that the reduction ratio, which is a stability determination parameter, is less than 1. On the other hand, an operation region with a small margin with respect to the reduction ratio 1 is excluded by a selected control rod (Selected Rods Insertion: SRI) or a stability restriction curve.

  The types of stability include channel stability with particular attention to the thermo-hydraulic stability of the highest output channel, core stability (fundamental mode stability), which is neutron flux oscillation with the entire core phase aligned, and core There is region stability (stability of higher-order modes) that is neutron flux oscillation that has an axis of symmetry in the circumferential direction and is 180 degrees out of phase. The sensitivity of each stability to the axial power distribution is more severe as the core stability is generally flat, and the channel stability and region stability are more severe as the lower peak distribution. In the core stability, the sensitivity to the axial power distribution differs from other stability. The core stability has a large effect of nuclear feedback, which has a large effect when the power peak is high at a high void fraction. This is because it appears as

  FIG. 63 is a graph showing a core stability analysis result of a BWR plant loaded with all 9 × 9 fuel (type A). Here, the stability analysis is performed using a frequency domain stability analysis code that models a BWR core including a coolant recirculation system.

  The core stability reduction ratio is the largest at the maximum output point on the minimum pump speed curve, and the reduction ratio is 0.70 in the core in which all the 9 × 9 fuel A types are loaded.

  FIG. 64 is a partially cutaway perspective view of a 9 × 9 fuel assembly (type B). FIG. 65 is a cross-sectional view showing a 9 × 9 fuel assembly (type B) together with control rods. In the 9 × 9 fuel assembly (type B), a water channel 52 for nine fuel rods is employed as a single-phase portion in the fuel assembly.

  As described above, in the BWR core, the deterioration of the limit output characteristic, the transient characteristic, and the stability is suppressed by various design improvements. On the other hand, light water reactor fuels are being burned up for the purpose of effectively using uranium resources and reducing the amount of spent fuel, and with this, the adoption of 10 rows and 10 columns (10 x 10) fuel is awaited. ing. At this time, the shape from the core axis power distribution, the lower strain distribution at the beginning of the cycle (inlet peak) to the upper strain distribution (outlet peak) from the middle to the end of the cycle appears as a larger peaking coefficient. It is in.

Non-patent document 2 describes a design example of high burnup 10-row 10-column fuel. In this example, 8 to 14 partial-length fuel rods are used. Patent Document 1 discloses an example in which one square water rod is employed with 10 × 10 fuel and two large-diameter water rods are employed.
JP 2005-134179 A JGM Andersen and two others, "Application of Advanced Thermal Hydraulic TRACG Model to Preserve Operating Margins in BWRs at Extended Power Uprate Conditions", Proceedings of the 2006 International Congress on Advances in Nuclear Power Plants (ICAPP '06), USA, ANS, 2006, p.1066 Nuclear Engineering International, UK, Nuclear Engineering International, September 2002

  The BWR scram characteristic generally decreases when the axial power distribution is strained upward (a maximum value appears near the core outlet). During a BWR rated power operation, if a transient change that increases the pressure occurs, for example, when the turbine trip or bypass valve is not operating, the reactor power will increase due to the disappearance of some voids in the core, and the thermal health of the fuel will The margin will decrease. That is, the possibility that the boiling state on the surface of the cladding tube transitions from nucleate boiling to film boiling increases. Therefore, it is detected that the turbine steam stop valve is closed, and a reactor scram signal is generated.

  However, in BWR, since the control rod 32 is inserted from the lower part of the core 3, the effect of negative reactivity injection by the scram appears rapidly in the lower part of the core 3, but it takes several seconds until the upper part of the core 3 is reached. A time delay occurs. For this reason, when the axial output distribution is strained upward, the margin for boiling transition tends to decrease, and ΔMCPR tends to increase. Since the negative void coefficient tends to increase at the end of the cycle, care must be taken when making a core operation plan so that the axial power distribution at the end of the cycle does not become extremely distorted.

  The BWR scram characteristics are also related to the insertion speed of the control rod 32. A plant having a control rod drive system with a high insertion speed is called a high-speed scrum plant. In the high-speed scrum plant, the insertion time during scrum at the average pressure of all cores at the rated pressure is 1.62 seconds or less when 75% of the total stroke is inserted. On the other hand, a plant having a control rod drive system with a slow insertion speed is conventionally called a scrum plant. In the conventional scrum plant, the average scrum time is 3.5 seconds or less with 90% insertion of the entire stroke.

  In the high-speed scrum plant, ΔMCPR at the time of transition is smaller than that of the conventional scrum plant even in the core design with the power distribution having the upper strain. In addition, the scrum characteristics of the improved BWR (ABWR) are intermediate characteristics between the high-speed scrum and the conventional scrum. In ABWR, the scram insertion time at the average pressure at the rated pressure is 1.44 seconds or less when 60% of all strokes are inserted and 2.80 seconds or less when 100% is inserted.

An abnormal transient change during operation that occurs in a BWR plant includes a transient change that hardly depends on the scram speed and the core cycle burnup, such as loss of feed water heating. Typical ΔMCPR when a generator load shut-off, bypass valve malfunction, and feed water heating loss occur in a high-speed scram plant, ABWR plant, or conventional scrum plant (end-of-cycle core) loaded with 9 × 9 fuel type A The results of the analysis are shown below.

  The operational limit MCPR (OLMCPR) is determined by adding the maximum value of ΔMCPR at the time of transition to the safety limit MCPR (SLMCPR, typically 1.07). Therefore, in the conventional scrum plant, OLMCPR can be reduced by improving the scrum characteristics. At the same time, when selecting a means to improve the scram characteristics, it is important to select a means that reduces ΔMCPR by about 0.1 when the generator load is shut off and the bypass valve is not operating. It can also be seen that doing this is meaningless in reducing OLMCPR.

  Note that ΔMCPR when the generator load is shut off and the bypass valve is not operating also depends on the water region (the cross-sectional area in the water rod) in the fuel assembly. The design of the water region is performed in consideration of mitigating ΔMCPR during a pressure rise transient, that is, reducing the negative void coefficient. In addition, it is also considered that positive reactivity is not input at the time of flow rate decrease transient or accident, that is, the void coefficient does not become positive. Furthermore, the improvement of the neutron economy by the efficient deceleration of neutrons, that is, the reduction of the fuel cycle cost is considered.

  In the high burnup fuel, a water region of about 7 to 10 times the cross-sectional area of the fuel rod may be employed. As a result, ΔMCPR in the conventional scrum plant has a fluctuation range of about 0.02, but there is still a demand for mitigating ΔMCPR by improving the scrum characteristics.

  As a means for improving the reactor scram characteristics, there is a method of shortening the effective core length. As a result, it can be expected that the effect of the scram appears immediately near the core exit. However, when the reactor core is shortened, it may be difficult to ensure the required cycle burnup, and the number of fuel replacement bodies tends to increase. Further, increasing the number of assemblies loaded on the short core, that is, increasing the core equivalent diameter, is not realistic because a large-scale modification of the plant is required.

  When high burnup 10x10 fuel is introduced to further improve the economics of the BWR core, the fuel pellet enrichment increases to achieve high burnup, and the nuclear feedback effect is large. Become. For this reason, there is a tendency that the margin of fuel thermal soundness during transition and the margin of nuclear thermal coupling (core stability, region stability) are reduced. Furthermore, the upper strain axial power distribution at the end of the cycle tends to have a larger peaking coefficient. Therefore, it is awaited to realize a core fuel that can achieve high burnup without reducing the soundness of the fuel / core.

  SUMMARY OF THE INVENTION Accordingly, an object of the present invention is to improve scram / transient characteristics and stability without reducing the economics of BWR.

  In order to achieve the above-described object, the present invention provides a fuel assembly in which a nuclear fuel material is placed in an effective fuel portion extending in the axial direction and loaded in a boiling water nuclear reactor, and a plane perpendicular to the axial direction extends. A fuel rod in which the nuclear fuel material is stored in at least a part of the fuel effective portion of a cylindrical cladding tube arranged at least at a part of a square lattice position of N rows and N columns, and a lower part that supports the fuel rod at the lower end A tie plate and an upper tie plate that supports at least a portion of the fuel rod at its upper end, and the fuel rod includes a full length fuel rod containing the nuclear fuel material over the entire fuel effective portion; A partially deficient fuel rod formed in a range in which the deficient region in which the nuclear fuel material is not stored extends from the upper end of the fuel effective portion downward by 1/6 or more of the fuel effective length that is the length of the fuel effective portion; Including the full length fuel rod Said the ratio of the volume of the defect area to the volume of the nuclear fuel material when placed in all positions of the tetragonal lattice positions is less than 3% and not more than 6%, and wherein.

The present invention also provides an N-row N-column square in a direction in which a plane perpendicular to the axial direction extends in a fuel assembly in which a nuclear fuel material is placed in an axially effective fuel portion and loaded into a boiling water reactor. A fuel rod in which the nuclear fuel material is housed in at least a part of the fuel effective portion of a cylindrical cladding tube arranged at least at a part of the lattice position; a lower tie plate that supports the fuel rod at a lower end; and the fuel An upper tie plate that supports at least a part of the rod at its upper end, and the fuel rod is a full-length fuel rod that contains the nuclear fuel material over the entire fuel effective portion, and the nuclear fuel material is contained in the fuel rod. The number of non-depleted regions formed in a range extending from the upper end of the fuel effective portion downward to 1/6 or more of the fuel effective length, which is the length of the fuel effective portion, between 1/5 and 1/3 of N ² A partially deficient fuel rod, and It is characterized by that.

The present invention also provides an N-row N-column square in a direction in which a plane perpendicular to the axial direction extends in a fuel assembly in which a nuclear fuel material is placed in an axially effective fuel portion and loaded into a boiling water reactor. A fuel rod in which the nuclear fuel material is housed in at least a part of the fuel effective portion of a cylindrical cladding tube arranged at least at a part of the lattice position; a lower tie plate that supports the fuel rod at a lower end; and the fuel An upper tie plate that supports at least a portion of the rod at its upper end, and the fuel rod contains the nuclear fuel material over the entire fuel effective portion and at least the lower end of the fuel effective portion and the fuel effective The enrichment is less than or equal to natural uranium in a range from a position distant from 1/12 of the effective fuel length, which is the length of the portion, to a position distant from the upper end of the effective fuel portion and 1/12 of the effective fuel length. Nuclei other than uranium Charge and full length fuel rods substances videos, enrichment N ² for defective areas of matches natural uranium following uranium is formed into 1/6 or extending the range of the fuel effective length downward from the upper end of the fuel effective portion Partly deficient fuel rods of 1/5 or more and 1/3 or less of the number of the deficient regions with respect to the volume of the nuclear fuel material when the full length fuel rods are arranged at all positions of the square lattice position. The volume ratio is 3% or more and 6% or less.

The present invention also provides an N-row N-column square in a direction in which a plane perpendicular to the axial direction extends in a fuel assembly in which a nuclear fuel material is placed in an axially effective fuel portion and loaded into a boiling water reactor. A fuel rod in which the nuclear fuel material is housed in at least a part of the fuel effective portion of a cylindrical cladding tube arranged at least at a part of the lattice position; a lower tie plate that supports the fuel rod at a lower end; and the fuel An upper tie plate that supports at least a portion of the rod at its upper end, and the fuel rod contains the nuclear fuel material over the entire fuel effective portion and at least the lower end of the fuel effective portion and the fuel effective The enrichment is less than or equal to natural uranium in a range from a position distant from 1/12 of the effective fuel length, which is the length of the portion, to a position distant from the upper end of the effective fuel portion and 1/12 of the effective fuel length. Nuclei other than uranium Charge and full length fuel rods substances videos, enrichment N ² for defective areas of matches natural uranium following uranium is formed into 1/6 or extending the range of the fuel effective length downward from the upper end of the fuel effective portion Of partially deficient fuel rods in the number of 1/5 or more and 1/3 or less.

The present invention also provides an N-row N-column square in a direction in which a plane perpendicular to the axial direction extends in a fuel assembly in which a nuclear fuel material is placed in an axially effective fuel portion and loaded into a boiling water reactor. A fuel rod in which the nuclear fuel material is housed in at least a part of the fuel effective portion of a cylindrical cladding tube arranged at least at a part of the lattice position; a lower tie plate that supports the fuel rod at a lower end; and the fuel An upper tie plate that supports at least a part of the rod at its upper end, and the fuel rod comprises a full-length fuel rod containing solid pellets of the nuclear fuel material over the entire fuel effective portion, and the fuel The solid pellets are accommodated in a first range extending from ½ to 5/6 of the effective fuel length, which is the length of the effective fuel portion, upward from the lower end of the effective portion, and from the upper end of the first range, The second range up to the top of the effective fuel section Including an upper hollow rods of 1/3 or more of the number of N ² that contains the hollow pellets with less than 1/4 of the inner diameter of the in the outer diameter of the actual pellet and an outer diameter the same as said solid pellets It is characterized by that.

The present invention also provides an N-row N-column square in a direction in which a plane perpendicular to the axial direction extends in a fuel assembly in which a nuclear fuel material is placed in an axially effective fuel portion and loaded into a boiling water reactor. A fuel rod in which the nuclear fuel material is housed in at least a part of the fuel effective portion of a cylindrical cladding tube arranged at least at a part of the lattice position; a lower tie plate that supports the fuel rod at a lower end; and the fuel An upper tie plate that supports at least a portion of the rod at its upper end; and the fuel rod includes a full length fuel rod containing a first solid pellet of the nuclear fuel material throughout the fuel active portion; The first solid pellets are housed in a first range extending from ½ to 5/6 of the effective fuel length, which is the length of the effective fuel portion, upward from the lower end of the effective fuel portion, From the upper end of the range to the upper end of the fuel effective part Including an upper small diameter fuel rods in the second 1/3 or more of the number of N ² that contains the solid pellets in a having an outer diameter of less than 90% of the outer diameter of the solid pellets within range of the first It is characterized by that.

The present invention also provides an N-row N-column square in a direction in which a plane perpendicular to the axial direction extends in a fuel assembly in which a nuclear fuel material is placed in an axially effective fuel portion and loaded into a boiling water reactor. A fuel rod in which the nuclear fuel material is housed in at least a part of the fuel effective portion of a cylindrical cladding tube arranged at least at a part of the lattice position; a lower tie plate that supports the fuel rod at a lower end; and the fuel An upper tie plate that supports at least a part of the rod at the upper end, and the fuel rod is ½ or more of the effective fuel length that is the length of the effective fuel portion upward from the lower end of the effective fuel portion. A first segment in which the nuclear fuel material is contained in a range extending 5/6 or less and helium is enclosed, and the nuclear fuel material is contained in a range connected to the upper end of the first segment and up to the upper end of the effective fuel portion. Said first seg Comprising an upper low pressurized fuel rods 1/3 or more the number of N ² encapsulated helium with 50% or less of the pressure of the cement in the initial helium filling pressure, and wherein the.

The present invention also provides an N-row N-column square in a direction in which a plane perpendicular to the axial direction extends in a fuel assembly in which a nuclear fuel material is placed in an axially effective fuel portion and loaded into a boiling water reactor. A fuel rod in which the nuclear fuel material is housed in at least a part of the fuel effective portion of a cylindrical cladding tube arranged at least at a part of the lattice position; a lower tie plate that supports the fuel rod at a lower end; and the fuel An upper tie plate that supports at least a part of the rod at the upper end, and the fuel rod is ½ or more of the effective fuel length that is the length of the effective fuel portion upward from the lower end of the effective fuel portion. The nuclear fuel material is contained in a first range extending 5/6 or less, and an average enrichment in a second range from the upper end of the range to the upper end of the effective fuel portion is 1 / of the average enrichment of the first range. 1.1 to reduce the nuclear fuel material Including 1/3 upper low-enriched fuel rods than in the number of meta-N ^2, and characterized in that.

The present invention also provides an N-row N-column square in a direction in which a plane perpendicular to the axial direction extends in a fuel assembly in which a nuclear fuel material is placed in an axially effective fuel portion and loaded into a boiling water reactor. A fuel rod in which the nuclear fuel material is housed in at least a part of the fuel effective portion of a cylindrical cladding tube arranged at least at a part of the lattice position; a lower tie plate that supports the fuel rod at a lower end; and the fuel An upper tie plate that supports at least a part of the rod at the upper end, and the fuel rod is ½ or more of the effective fuel length that is the length of the effective fuel portion upward from the lower end of the effective fuel portion. The first pellet of the nuclear fuel material is accommodated in a first range extending 5/6 or less, and the average density is in the second range from the upper end of the range to the upper end of the fuel effective part. 1 / 1.1 or less of the nuclear fuel Including a second one-third or more of the number of N ² that contains the pellet of the upper low-density fuel rod quality, and characterized in that.

The present invention also provides an N-row N-column square in a direction in which a plane perpendicular to the axial direction extends in a fuel assembly in which a nuclear fuel material is placed in an axially effective fuel portion and loaded into a boiling water reactor. A fuel rod in which the nuclear fuel material is housed in at least a part of the fuel effective portion of a cylindrical cladding tube arranged at least at a part of the lattice position; a lower tie plate that supports the fuel rod at a lower end; and the fuel An upper tie plate that supports at least a part of the rod at the upper end, and the fuel rod is ½ or more of the effective fuel length that is the length of the effective fuel portion upward from the lower end of the effective fuel portion. The nuclear fuel material is contained in a first range extending 5/6 or less, and an average gadolinia concentration in a second range from the upper end of the range to the upper end of the fuel effective part is 1 / of the average gadolinia concentration of the first range. 1.1 or less Including a Kikaku fuel material and 1/3 or more of the number of N ² of matches gadolinia upper low-enriched fuel rods, and characterized in that.

  According to the present invention, scram / transient characteristics and stability can be improved without reducing the economics of BWR.

  An embodiment of a fuel assembly according to the present invention will be described with reference to the drawings. In addition, the same code | symbol is attached | subjected to the same or similar structure, and the overlapping description is abbreviate | omitted.

[First Embodiment]
FIG. 4 is a cross-sectional view taken along arrows Z4-Z4 in FIGS. 6 and 7, schematically showing a cross-section in the first embodiment of the fuel assembly according to the present invention. FIG. 5 is a cross-sectional view taken along arrows Z5-Z5 in FIGS. 6 and 7, schematically showing a cross-section in the first embodiment of the fuel assembly according to the present invention. 6 is a vertical cross-sectional view taken along arrows X6-X6 in FIGS. 7 is a vertical sectional view taken along arrows X7-X7 in FIGS.

  The fuel assembly of the present embodiment is a fuel assembly 30 loaded in the core 3 of the boiling water nuclear power plant shown in FIGS. 52 and 54, for example. The fuel assembly 30 includes a fuel rod 51, a lower tie plate 62, and an upper tie plate 61.

  The fuel rods 51 are arranged in a part of a 10 × 10 square lattice position in a plane perpendicular to the axial direction, that is, in a horizontal plane when the core 3 is loaded. The fuel rod 51 is obtained by accommodating nuclear fuel material as pellets 54 of uranium oxide in a cladding tube 55 whose upper and lower ends are sealed with end plugs. The outer diameter of the cladding tube 55 is about 10.2 mm, and the diameter of the loaded pellet 54 is about 8.8 mm.

  The fuel assembly 30 has a water channel 52. The water channel 52 is (5,5), (5,6), (5,7), (6,5), (6,6), (6,7), (7,5), (7, The fuel rods 51 are arranged at positions corresponding to nine fuel rods 51 represented by 6) and (7, 7). Here, (i, j) represents the row number i and the column number j for each of the above-mentioned square lattice positions.

  The lower tie plate 62 supports the lower end of the fuel rod 51. The upper tie plate 61 supports the upper ends of some fuel rods 51. Further, the fuel assembly 30 has, for example, eight spacers 63 arranged at intervals in the axial direction. Both the lower tie plate 62 and the upper tie plate 61 have the function of arranging the fuel rods 51 at the square lattice positions and maintaining the distance between the fuel rods 51 and the water channels 52.

  The fuel rods 51 are arranged at 91 points out of 100 square lattice positions in 10 rows and 10 columns. A range in the axial direction in which the pellets 54 are stored in any one of the fuel rods 51 is referred to as a fuel effective portion, and the length is referred to as a fuel effective length A. The effective fuel length is about 3.7 m.

  The fuel rod 51 of the present embodiment has a full-length fuel rod 71 that contains pellets 54 over the entire effective fuel portion, and partially deficient fuel in which deficient regions 91 and 92 that do not contain pellets 54 are formed. And bars 72 and 82. The total length of the full length fuel rod 71 is about 4 m. The partially deficient fuel rods 72 and 82 contain pellets 54 in a range B extending about 3.0 m upward from the lower end.

  In the present embodiment, there are two types of partially missing fuel rods 72, 82, full length partially missing fuel rod 72 and short length partially missing fuel rod 81.

  The full-length partially deficient fuel rod 72 is a fuel rod 51 having the same length as the full-length fuel rod 71 as a whole, but in which a pellet 54 is accommodated in a range B extending about 3.0 m upward from the lower end. Accordingly, the upper end of the full length partially missing fuel rod 72 is supported by the upper tie plate 61, similarly to the full length fuel rod 71. Partially missing fuel rods 72 are (1,4), (1,7), (4,1), (7,1), (10,4), (10,7), (4,10) And (7, 10). The deficient region 92 of the full length partially deficient fuel rod 72 is a plenum.

  The short partly deficient fuel rod 81 is a fuel rod 51 shorter than the full length fuel rod 71 as a whole, and the pellets 54 are stored in a range B extending about 3.0 m upward from the lower end. Therefore, the short partially missing fuel rod 81 is not supported by the upper tie plate 61. The short partially missing fuel rods 81 are (1,5), (1,6), (5,1), (6,1), (10,5), (10,6), (5,10) , (6,10), (4,5), (4,7), (5,4), (5,8), (7,4), (7,8), (8,5) and ( It is arranged at each of the 16 locations represented by 8, 7).

The fuel assembly 30 has a volume of pellets 54 to be loaded, that is, a volume obtained by multiplying the effective fuel length by the pellet outer diameter is about 1.95 × 10 ⁴ cm ³ , and 9 × 9 fuel (type A) Is approximately 3% higher than the loaded fuel volume. Therefore, higher burnup can be achieved than with 9 × 9 fuel (A type).

  By arranging eight part-length fuel rods around the water channel 52, the water to uranium ratio in the central region of the assembly is increased. For this reason, the reduction effect of a negative void coefficient becomes large. In addition, by arranging a total of eight full-length partially missing fuel rods 72 and eight short partially missing fuel rods 81 on the four sides of the outer periphery of the assembly closest to the bypass region, an effect of reducing the void coefficient can be obtained. Enlarged. Further, since the full length partially deficient fuel rod 72 and the short length partially deficient fuel rod 81 are arranged avoiding the assembly corner portion, the influence on the LPRM 33 which is a part of the neutron instrumentation system is minimized. It has become.

  Further, in the present embodiment, when the effective lengths of all the fuel rods 51 are the same, the defect regions 91 and 92 in which the pellets are not loaded are 5/6 or more and 1 or less of the effective fuel length A. The ratio of the number of partially missing fuel rods 72 and 81 existing in the range and the ratio of the fuel rods to the total cross-sectional area when the fuel rods are arranged at all lattice positions are both 24%. Further, the ratio of the volume of the defect regions 91 and 92 to the nuclear fuel material loading volume when the full length fuel rods 71 are arranged at all lattice positions is about 5%.

  FIG. 8 is a graph showing an example of the core average axial power distribution at the beginning and the end of the cycle of the core loaded with the fuel assembly in the present embodiment. At the end of the cycle, the control rod is fully pulled out.

  In the present embodiment, the nuclear fuel material at the upper part of the fuel assembly 30 is significantly reduced, so that the upper part of the core loaded with 9 × 9 fuel (type A) is compared with the example of the axial power distribution at the end of the cycle. The power generation rate at the (core exit) has decreased by about 20%.

  9 to 11 are graphs showing transient analysis results when the turbine trip / bypass valve does not operate in the end-cycle core of the conventional scrum plant loaded with the fuel assembly of the present embodiment. TRACG code is used as the analysis code. 10 to 11 also show values for a 9 × 9 fuel (A type) loaded core for comparison.

  In the core 3 loaded with the fuel assembly 30 of the present embodiment, the initial axial power distribution is not strained upward, and the water-to-uranium ratio at the top of the core is increased, that is, the negative void coefficient is reduced. ing. For this reason, the negative reactivity due to the scram action greatly affects the cancellation of the positive reactivity input due to void disappearance. As a result, ΔMCPR (0.17) at the time of transition is 0.08 smaller than ΔMCPR (0.25) in the 9 × 9 fuel (A type) loaded core.

  FIG. 12 is a graph showing the core stability reduction ratio of the core loaded with the fuel assembly of the present embodiment. For comparison, FIG. 12 also shows values in a 9 × 9 fuel (A type) loaded core.

  The core stability reduction ratio of the core 3 loaded with the fuel assembly 30 of the present embodiment is 9 × 9 due to the effect of reducing the negative void coefficient and the effect of reducing the pressure loss by using the partial length fuel rods. Compared with the fuel (A type) loaded core, the maximum value (value at the minimum pump speed maximum output point) is improved by about 0.10.

  As described above, in this embodiment, the axial output distribution at the end of the cycle does not have an extreme peaking coefficient at the upper part of the fuel assembly 30, and the fuel cladding tube-coolant heat at the time when the fuel temperature rises. The transfer characteristic can be made gentle. Therefore, the fuel assembly according to the present embodiment can improve the scram / transient characteristics and stability without reducing the economics of BWR.

[Second Embodiment]
FIG. 13 is a cross-sectional view taken in the direction of arrows Z13-Z13 in FIGS. 15 and 16, schematically showing a cross-section in the second embodiment of the fuel assembly according to the present invention. FIG. 14 is a Z14-Z14 arrow cross-sectional view in FIGS. 15 and 16 schematically showing a cross-section in the second embodiment of the fuel assembly according to the present invention. 15 is a longitudinal sectional view taken along arrows X15-X15 in FIG. 13 and FIG. 16 is a vertical cross-sectional view taken along arrows X16-X16 in FIG. 13 and FIG.

  The fuel assembly 30 of the present embodiment is different from the fuel assembly of the first embodiment in the type and arrangement of the fuel rods 51.

  In the present embodiment, there are three types of fuel rods 51, a full length fuel rod 71, a first short partially missing fuel rod 81, and a second short partially missing fuel rod 82. The first short partly deficient fuel rod 81 is a fuel rod 51 shorter than the full length fuel rod 71 as a whole, and the pellet 54 is accommodated in a range B extending about 3.0 m upward from the lower end. The second short partly deficient fuel rod 82 is a fuel rod 51 shorter than the first short partly deficient fuel rod 81 as a whole, and the pellets 54 are in a range C extending about 2.0 m upward from the lower end. It is stored. The first and second short partly missing fuel rods 81 and 82 are not supported by the upper tie plate 61.

  In this fuel assembly 30, two of the 16 short partly deficient fuel rods 81 arranged in the peripheral portion of the fuel assembly 30 of the first embodiment, which are farthest from the center of the control rod 32, The full length fuel rod 71 is replaced. In addition, of the eight short partially missing fuel rods 81 arranged around the water channel in the fuel assembly 30 according to the first embodiment, two of the shortest fuel rods 81 farthest from the center of the control rod 32 are the full length fuel rods 71. And the remaining six second short partly missing fuel rods 82 are replaced.

  That is, the first short partly deficient fuel rod 81 has (1,4), (1,5), (1,6), (1,7), (4,1), (5,1), (6,1), (7,1), (10,4), (10,5), (10,6), (4,10), (5,10) and (6,10) It is arranged at each of 14 locations. The second short partially missing fuel rods 82 are (4, 5), (4, 7), (5, 4), (5, 8), (7, 4) and (8, 5). It is arrange | positioned at each of six places represented.

  In the fuel assembly 30 of the present embodiment, the defect region 91 exists in the range from 1/2 to 6/6 of the effective fuel length A. Further, the number of fuel rods 51 having the defective region 91 in 1/2 to 2/3 of the effective fuel length A and the number of fuel rods 51 having the defective region 91 in 2/3 to 5/6 of the effective fuel length A The total ratio is 20%, and the ratio of the number of missing fuel rods at 1/2 to 2/3 is 6%. The ratio of the cross-sectional area of the defect region 91 at the effective fuel length A of 5/6 to 6/6 to the total cross-sectional area of the fuel rods when the fuel rods are arranged at all lattice positions is 20%. The ratio of the volume of the defect region 91 to the loading volume of the nuclear fuel material when the full length fuel rods 71 are arranged at all lattice positions is about 5%.

  Such a fuel assembly 30 can also improve the scram characteristics after alleviating the distortion of the neutron flux distribution in the cross section of the fuel assembly 30 due to the eccentricity of the water channel 52.

  FIG. 17 is a graph showing an example of the core average axial power distribution at the beginning and end of the cycle of the core loaded with the fuel assembly in the present embodiment. At the end of the cycle, the control rod is fully pulled out.

  In the present embodiment, the length of the second short partially missing fuel rod 82 around the water channel 52 is about 2.0 m, so that the missing area 91 is enlarged. For this reason, the decrease in the upper part of the axial output distribution can be made smoother than in the first embodiment.

ΔMCPR at the time of transition is improved by 0.08 over 9 × 9 fuel (A type), and the core stability reduction ratio is improved by about 0.10 at the maximum value compared with 9 × 9 fuel (A type). . The volume of fuel loaded (approximately 1.93 × 10 ⁴ cm ³ ) is approximately 2% higher than 9 × 9 fuel (type A). Therefore, the fuel assembly according to the present embodiment can improve the scram / transient characteristics and stability without reducing the economics of BWR.

[Third Embodiment]
FIG. 18 is a cross-sectional view taken along arrows Z18-Z18 in FIGS. 20 and 21, schematically showing a cross-section in the third embodiment of the fuel assembly according to the present invention. FIG. 19 is a cross-sectional view taken along arrows Z19-Z19 in FIGS. 20 and 21, schematically showing a cross-section in the third embodiment of the fuel assembly according to the present invention. 20 is a vertical cross-sectional view taken along arrows X20-X20 in FIGS. 21 is a vertical cross-sectional view taken along arrows X21-X21 in FIGS.

  The fuel assembly 30 of the present embodiment differs from the fuel assembly 30 of the second embodiment in the position of the water channel 52 and the arrangement of the fuel rods 51.

  The water channel 52 is (4, 4), (4, 5), (4, 6), (5, 4), (5, 5) closer to the center of the control rod 32 than in the second embodiment. , (5, 6), (6, 4), (6, 5) and (6, 6). Further, the fuel rod 51 is disposed for the purpose of alleviating the distortion of the neutron flux distribution in the fuel assembly 30 due to the eccentricity of the water channel 52. The first short partly deficient fuel rod 81 has (1,5), (1,6), (1,7), (5,1), (6,1), (7,1), (10 , 4), (10, 5), (10, 6), (10, 7), (4, 10), (5, 10), (6, 10) and (7, 10) 14 It is arranged in each of the places. The second short partly missing fuel rod 82 is represented by (3, 6), (4, 7), (6, 3), (6, 7), (7, 4) and (7, 6). It is arranged at each of the six locations.

  In the fuel assembly 30 of the present embodiment, the defect region 91 exists in the range from 1/2 to 6/6 of the effective fuel length A. Further, the number of fuel rods 51 having the defective region 91 in 1/2 to 2/3 of the effective fuel length A and the number of fuel rods 51 having the defective region 91 in 2/3 to 5/6 of the effective fuel length A The total ratio is 20%, and the ratio of the number of missing fuel rods at 1/2 to 2/3 is 6%. The ratio of the cross-sectional area of the defect region 91 at the effective fuel length A of 5/6 to 6/6 to the total cross-sectional area of the fuel rods when the fuel rods are arranged at all lattice positions is 20%. The ratio of the volume of the defect region 91 to the loading volume of the nuclear fuel material when the full length fuel rods 71 are arranged at all lattice positions is about 5%.

  By setting the pellet loading length of the second short partly deficient fuel rod 82 around the water channel 52 to about 2.0 m, the expansion of the fuel deficient region is stepwise. For this reason, the decrease in the upper part of the axial output distribution is smoother than in the first embodiment.

ΔMCPR at the time of transition is improved by 0.08 over 9 × 9 fuel (A type), and the core stability reduction ratio is improved by about 0.10 at the maximum value compared with 9 × 9 fuel (A type). . Moreover, the volume of the pellets to be loaded (about 1.93 × 10 ⁴ cm ³ ) is about 2% higher than the high burnup 9 × 9 fuel (A type). Therefore, the fuel assembly according to the present embodiment can improve the scram / transient characteristics and stability without reducing the economics of BWR.

[Fourth Embodiment]
FIG. 22 is a cross-sectional view taken along arrows Z22-Z22 in FIGS. 24 and 25, schematically showing a cross-section in the fourth embodiment of the fuel assembly according to the present invention. FIG. 23 is a cross-sectional view taken along arrows Z23-Z23 in FIGS. 24 and 25, schematically showing a cross-section in the fourth embodiment of the fuel assembly according to the present invention. 24 is a longitudinal sectional view taken along arrows X24-X24 in FIGS. 22 and 23. FIG. 25 is a vertical sectional view taken along arrows X24-X24 in FIG. 22 and FIG.

  The fuel assembly 30 of the present embodiment uses a water rod 53 instead of the water channel 52 in the fuel assembly 30 of the first to third embodiments. The water rod 53 occupying the position of the four fuel rods 51 is located at the positions (4, 6), (4, 7), (5, 6) and (5, 7) in the approximate center of the 10 × 10 lattice. , And (6, 4), (6, 5), (7, 4) and (7, 5). The maximum outer diameter of the water rod 53 is about 25 mm.

  The fuel assembly 30 includes sixteen first short partially missing fuel rods 81 and six second short partially missing fuel rods 82 in addition to the 70 full length fuel rods 71. The pellet loading length of the first short partially missing fuel rod 81 is about 3.0 m, and the pellet loading length of the second short partially missing fuel rod 82 is about 2.0 m. By using these short partly deficient fuel rods 81 and 82, the expansion of the deficient region 91 is gradual.

  The first short partly deficient fuel rod 81 has (1, 4), (1, 5), (1, 6), (1, 7), (4, 1), (5, 1) on the outer periphery. , (6,1), (7,1), (10,4), (10,5), (10,6), (10,7), (4,10), (5,10), ( 6, 10) and (7, 10). The second short partly deficient fuel rods 82 are (4, 5), (5, 4), (5, 5), (6, 6), (6, 7) and (7) around the water rod 53. , 6).

  By disposing the six second short partly deficient fuel rods 82 around the water rod 53, the water-to-uranium ratio in the central region of the cross section of the fuel assembly 30 is increased. The effect of reducing the coefficient is increased. By disposing the first short partly deficient fuel rods 81 on the four sides of the outer periphery of the fuel assembly 30 closest to the bypass region, the void coefficient reduction effect is also expanded. Since the short partly deficient fuel rods 81 and 82 are arranged so as to avoid the corner portion of the fuel assembly 30, the influence of the neutron instrumentation system (LPRM33) is minimized.

  In the fuel assembly 30 of the present embodiment, the defect region 91 exists in the range from 1/2 to 6/6 of the effective fuel length A. Further, the number of fuel rods 51 having the defective region 91 in 1/2 to 2/3 of the effective fuel length A and the number of fuel rods 51 having the defective region 91 in 2/3 to 5/6 of the effective fuel length A The total ratio is 22%, and the ratio of the number of missing fuel rods at 1/2 to 2/3 is 6%. The ratio of the cross-sectional area of the defect region 91 at the effective fuel length A of 5/6 to 6/6 to the total cross-sectional area of the fuel rods when the fuel rods are arranged at all lattice positions is 22%. The ratio of the volume of the defect region 91 to the loading volume of the nuclear fuel material when the full length fuel rods 71 are arranged at all lattice positions is about 6%.

The ΔMCPR at the time of transition is improved by 0.07 over 9 × 9 fuel (A type), and the maximum value of the core stability reduction ratio is improved by 0.07 over 9 × 9 fuel (A type). Moreover, the volume of the pellets to be loaded (about 1.94 × 10 ⁴ cm ³ ) is about 2% higher than the high burnup 9 × 9 fuel (type A). Therefore, the fuel assembly according to the present embodiment can improve the scram / transient characteristics and stability without reducing the economics of BWR.

[Fifth Embodiment]
FIG. 26 is a cross-sectional view taken along arrows Z26-Z26 in FIGS. 24 and 25, schematically showing a cross-section in the fifth embodiment of the fuel assembly according to the present invention. FIG. 27 is a cross-sectional view taken along arrows Z27-Z27 in FIGS. 24 and 25, schematically showing a cross-section in the fifth embodiment of the fuel assembly according to the present invention. 28 is a longitudinal sectional view taken along arrows X28-X28 in FIGS. 29 is a vertical cross-sectional view taken along arrows X29-X29 in FIGS. 22 and 23. FIG.

  Similar to the fuel assembly 30 of the first embodiment, the fuel assembly 30 of the present embodiment has a water channel 52 and a short partially missing fuel rod 81 having a pellet loading length of about 3.0 m. Yes. The water channel 52 is (5,5), (5,6), (5,7), (6,5), (6,6), (6,7), (7,5), (7, 6) and (7, 7). The short partially missing fuel rods 81 are (1,5), (1,6), (5,1), (6,1), (10,5), (10,6), (5,10) , (6, 10), (4, 6), (6, 4), (6, 8) and (8, 6).

  35 remaining full length fuel rods 71 and 40 upper hollow fuel rods 73 are disposed at the remaining 79 square lattice positions where the water channel 52 or the short partially missing fuel rod 81 is not disposed. The upper hollow fuel rods 73 are arranged so as not to be adjacent to each other. In the upper hollow fuel rod 73, the same solid pellet 54 is loaded at the same position as the pellet loading position of the short partially missing fuel rod 81, and the hollow pellet 56 is loaded above the same.

  When the hollow pellet 56 is loaded in the upper region of the fuel assembly 30, the water to uranium ratio in this region increases, so the negative void coefficient decreases. This effect is combined with the effect of avoiding the upward distortion of the axial output distribution at the end of the cycle by adopting the 12 short partly deficient fuel rods 81, thereby improving the transient characteristics when the void is reduced. Note that the reduction of the negative void coefficient acts to improve the core stability.

ΔMCPR at the time of transition is improved by 0.05 over 9 × 9 fuel (A type), and the core stability reduction ratio is improved by about 0.07 from the 9 × 9 fuel (A type). . The volume of fuel loaded (about 1.99 × 10 ⁴ cm ³ ) is about 5% higher than 9 × 9 fuel (type A). Therefore, the fuel assembly according to the present embodiment can improve the scram / transient characteristics and stability without reducing the economics of BWR.

[Sixth Embodiment]
FIG. 30 is a cross-sectional view taken along arrows Z30-Z30 in FIGS. 32 and 33, schematically showing a cross-section in the sixth embodiment of the fuel assembly according to the present invention. FIG. 31 is a Z31-Z31 arrow cross-sectional view in FIGS. 32 and 33 schematically showing a cross-section in the sixth embodiment of the fuel assembly according to the present invention. 32 is a vertical cross-sectional view taken along arrows X32-X32 in FIGS. 30 and 31. FIG. 33 is a vertical cross-sectional view taken along arrows X33-X33 in FIGS. 30 and 31. FIG.

  In the fuel assembly 30 of the present embodiment, the upper hollow fuel rod 73 of the fuel assembly 30 of the fifth embodiment is replaced with an upper small-diameter fuel rod 74. The upper small-diameter fuel rod 74 is loaded with a small-diameter pellet 57 instead of the hollow pellet 56 loaded in the upper hollow fuel rod 73. The small-diameter pellet 57 has an outer diameter of about 8 mm.

  FIG. 34 is a graph showing the MCPR analysis result of the core loaded with the fuel assembly of the present embodiment. FIG. 34 also shows the analysis result of the core loaded with 9 × 9 fuel (A type) for comparison.

  When the diameter of the pellet is reduced, the gap (gap) between the outer surface of the pellet and the inner surface of the cladding tube 55 increases, so that the gap heat transfer coefficient tends to decrease. As a result, when the positive reactivity is applied due to the disappearance of the void and the fuel rod temperature rises, the rate of increase of the heat flux from the cladding surface to the coolant becomes gradual. Combined with this effect and the effect of avoiding the upward distortion of the axial output distribution at the end of the cycle by adopting the 12 short partially missing fuel rods 81, ΔMCPR (0.20) at the time of transition is conventionally It becomes smaller than 9x9 fuel A type (0.25). Note that the decrease in the gap heat transfer coefficient acts to improve the core stability.

ΔMCPR at the time of transition is improved by 0.05 over 9 × 9 fuel (A type), and the core stability reduction ratio is improved by about 0.07 from the 9 × 9 fuel (A type). . The volume of fuel loaded (approximately 1.97 × 10 ⁴ cm ³ ) is approximately 4% higher than 9 × 9 fuel (type A). Therefore, the fuel assembly according to the present embodiment can improve the scram / transient characteristics and stability without reducing the economics of BWR.

[Seventh Embodiment]
FIG. 35 is a Z35-Z35 arrow cross-sectional view in FIGS. 37 and 38 schematically showing a cross-section in the seventh embodiment of the fuel assembly according to the present invention. FIG. 36 is a cross-sectional view taken along arrows Z36-Z36 in FIGS. 37 and 38, schematically showing a cross-section in the seventh embodiment of the fuel assembly according to the present invention. 37 is a longitudinal sectional view taken along arrows X37-X37 in FIGS. 35 and 36. FIG. 38 is a vertical sectional view taken along arrows X38-X38 in FIGS. 35 and 36. FIG.

  In the fuel assembly 30 of the present embodiment, the upper hollow fuel rod 73 of the fuel assembly of the fifth embodiment is replaced with a fuel rod in which the first and second segments 83 and 84 are coupled.

  FIG. 39 is a partially enlarged longitudinal sectional view of a segment in the present embodiment.

  The first and second segments 83 and 84 are obtained by loading a solid pellet 54 in the cladding tube 55, respectively. The first segment 83 and the second segment 84 are coupled by a segment coupling unit 85. A plenum spring 86 is pressed against the upper end of the pellet 54. A plenum 93 is formed by the plenum spring 86. Further, helium is enclosed in the first segment 83 at an initial enclosure pressure of 7 atm. The second segment 84 is filled with helium at an initial enclosure pressure of 3 atmospheres.

  When the initial helium filling pressure of the fuel rod is reduced, the gap heat transfer coefficient tends to decrease. As a result, when a positive reactivity is applied due to the disappearance of the void and the fuel rod temperature rises, the rate of increase of the heat flux from the cladding surface to the coolant becomes gentle, and the thermal integrity of the cladding 55 is reduced. The margin increases. Also, a plenum 93 is provided at the upper end of the lower first segment 83 to mitigate upper peaking due to a transient flow of neutron flux in the axial upper part.

  ΔMCPR at the time of transition is improved by 0.05 over 9 × 9 fuel (A type), and the core stability reduction ratio is improved by about 0.07 from the 9 × 9 fuel (A type). . The volume of fuel loaded is about 5% higher than 9 × 9 fuel (type A). Therefore, the fuel assembly according to the present embodiment can improve the scram / transient characteristics and stability without reducing the economics of BWR.

[Eighth Embodiment]
FIG. 40 is a Z40-Z40 arrow cross-sectional view in FIGS. 42 and 43 schematically showing a cross-section in the eighth embodiment of the fuel assembly according to the present invention. FIG. 41 is a cross-sectional view taken along arrows Z41-Z41 in FIGS. 42 and 43, schematically showing a cross-section in the eighth embodiment of the fuel assembly according to the present invention. 42 is a longitudinal sectional view taken along arrows X42-X42 in FIGS. 40 and 41. FIG. 43 is a longitudinal sectional view taken along arrows X43-X43 in FIGS. 40 and 41. FIG.

  In the fuel assembly 30 of the present embodiment, the full length fuel rod 71 in the fuel assembly of the sixth embodiment is replaced with an upper low enriched fuel rod 75. The upper low enriched fuel rod 75 is obtained by loading the low enriched pellet 58 at the same position as the position where the small diameter pellet 57 is loaded. The average concentration in the axial direction of the low concentration pellet 58 is about 3.6%, and the average concentration in the axial direction of the pellet 54 loaded below the position where the low concentration pellet 58 is loaded is 4%. . Since the average enrichment at the upper part of the upper low enriched fuel rod 75 is lower than the lower part, it is possible to avoid the axial output distribution from becoming an extreme upper peak at the end of the cycle.

ΔMCPR at the time of transition is improved by 0.08 over 9 × 9 fuel (A type), and the core stability reduction ratio is improved by about 0.10 at the maximum value compared with 9 × 9 fuel (A type). . The volume of fuel loaded (approximately 1.92 × 10 ⁴ cm ³ ) is approximately 1% higher than 9 × 9 fuel (type A). Therefore, the fuel assembly according to the present embodiment can improve the scram / transient characteristics and stability without reducing the economics of BWR.

  Furthermore, by taking the maximum pellet enrichment at about 3.0 m from the lower end to be larger than 5%, it is possible to extend the average burnup of the fuel assembly 30. In addition, by increasing the average pellet density at about 3.0 m from the lower end of the effective fuel portion to about 1.1 times or more than the average pellet density at about 3.0 m to about 3.7 m, at the end of the cycle The effect of avoiding that the axial output distribution becomes an extreme upper peak is obtained. The concentration of gadolinia, which is a neutron poison added at the time of pellet production, is about 1. for the average concentration at about 3.0 m from the lower end of the effective fuel portion than for the average concentration at about 3.0 m to about 3.7 m. By making it larger by a factor of 1 or more, the occurrence of an upper peak at the end of the cycle can be avoided.

[Ninth Embodiment]
FIG. 44 is a cross-sectional view taken along arrows Z44-Z44 in FIGS. 46 and 47, schematically showing a cross-section in the eighth embodiment of the fuel assembly according to the present invention. FIG. 45 is a cross-sectional view taken along arrows Z45-Z45 in FIGS. 46 and 47, schematically showing a cross-section in the eighth embodiment of the fuel assembly according to the present invention. 46 is a longitudinal sectional view taken along arrows X46-X46 in FIGS. 44 and 45. FIG. 47 is a longitudinal sectional view taken along arrows X47-X47 in FIGS. 44 and 45. FIG.

  The fuel assembly 30 of the present embodiment is obtained by replacing the fuel rods 72 having a partially missing full length of the fuel assembly 30 of the first embodiment with fuel rods 76 with an upper solid moderator. The upper solid moderator-containing fuel rod 76 is obtained by storing the solid moderator 59 in the deficient region 92 of the full length partially deficient fuel rod 72 in the first embodiment. The solid moderator 59 is, for example, zirconium hydride, zirconium deuteride, yttrium hydride, cerium hydride, or the like.

  In such a fuel assembly 30, the negative void coefficient in the upper region of the fuel assembly can be reduced by disposing the solid moderator in the upper region. As a result, the adoption of 16 short partly deficient fuel rods 81 combined with the effect of avoiding the upward distortion of the axial output distribution at the end of the cycle improves the transient characteristics and stability at the time of void reduction. Is done. A similar effect can be obtained by connecting an alloy rod containing a solid moderator to the short partly deficient fuel rod 81.

  ΔMCPR at the time of transition is improved by 0.08 over 9 × 9 fuel (A type), and the core stability reduction ratio is improved by about 0.10 at the maximum value compared with 9 × 9 fuel (A type). . The volume of fuel loaded is about 3% higher than 9 × 9 fuel (type A). Therefore, the fuel assembly according to the present embodiment can improve the scram / transient characteristics and stability without reducing the economics of BWR.

  Further, the defect region 92 from about 3.0 m to about 3.7 m from the lower end of the fuel effective portion may be loaded with natural uranium fuel pellets or deteriorated uranium fuel pellets in addition to the method of using as a plenum portion. This is also possible, and the relaxation effect of the upper peak can be obtained.

[Tenth embodiment]
FIG. 48 is a cross-sectional view taken along arrows Z48-Z48 in FIGS. 50 and 51, schematically showing a cross-section in the eighth embodiment of the fuel assembly according to the present invention. FIG. 49 is a cross-sectional view taken along arrows Z49-Z49 in FIGS. 50 and 51, schematically showing a cross-section in the eighth embodiment of the fuel assembly according to the present invention. 50 is a vertical cross-sectional view taken along arrows X50-X50 in FIGS. 48 and 49. FIG. 51 is a longitudinal sectional view taken along arrows X51-X51 in FIGS. 48 and 49. FIG.

  The fuel assembly 30 of the present embodiment is for a high-speed scrum plant or ABWR plant. This fuel assembly 30 is obtained by replacing the full-length partially defective fuel rod 72 disposed on the outer peripheral portion of the fuel assembly 30 of the first embodiment with a full-length fuel rod 71. That is, of the partially defective fuel rods arranged on the outer peripheral portion of the fuel assembly 30 of the first embodiment, two of the fuel effective portions of the total length of partially defective fuel rods 72 are two per side. The pellets 54 are loaded in the range of about 3.0 m to about 3.7 m from the lower end. Thereby, the volume of the loaded fuel exceeds the high burnup 9 × 9 fuel (A type) by about 4%.

  Moreover, ΔMCPR at the time of transition is improved by 0.15 from 9 × 9 fuel (A type), and the maximum stability reduction ratio is about 0.03 improved from 9 × 9 fuel (A type). Is done. Therefore, the fuel assembly according to the present embodiment can improve the scram / transient characteristics and stability without reducing the economic efficiency of the high-speed scrum plant or ABWR plant.

[Other embodiments]
In addition, this invention is not limited to each above-mentioned embodiment, It can implement with various forms. Moreover, it can also implement combining the characteristic of each embodiment.

  FIG. 66 is a table summarizing the features of each embodiment.

  The fuel assemblies 30 of the above-described embodiments are based on the BWR fuel design and manufacturing technology cultivated so far, such as short partly missing fuel rods 81 (partial length fuel rods), large diameter water rods, and the like. On top of this, it improves the scram / transient characteristics and stability without reducing the economics of the BWR. Also, these fuel assemblies 30 employ partial-length fuel rods of an appropriate number, length and arrangement, or use pellets 54 loaded with appropriate diameter, U235 enrichment, density and gadolinia concentration. Yes. As a result, the axial output distribution at the end of the cycle does not have an extreme peaking coefficient at the upper part of the assembly, and the fuel cladding-coolant heat transfer characteristic when the fuel temperature rises becomes gentle. .

  When the number of partially missing fuel rods 71, 81, 82 is increased, it is possible to easily avoid that the power distribution in the axial direction of the fuel assembly has a peak near the outlet when the core operation plan is formulated. In this case, the negative reactivity is effectively input during the scram operation, and the MCPR drop width (ΔMCPR) at the time of transition can be reduced.

  In general, the scram characteristic is improved when the axial power distribution is under strain. Therefore, setting the loading range of the nuclear fuel material in the fuel rod 51 to less than half of the maximum effective length does not lead to the improvement of the scrum characteristic. In addition, it is disadvantageous from the viewpoint of securing the fuel pellet loading amount to employ many fuel rods 51 having a short nuclear fuel material loading range.

  FIG. 3 is a graph showing an example of the relationship between the amount of nuclear fuel material loaded and the number of fuel rods for each length of loaded nuclear fuel material. This figure is an example of a 10 × 10 fuel assembly in which a water area for nine fuel rods is secured by one water channel and the effective fuel length A is about 3.71 m. The outer diameter of the cladding tube is about 10.2 mm, and the pellet diameter is about 8.8 mm. The fuel pellet loading amount is expressed as a ratio to the pellet loading volume of 9 × 9 fuel (A type). The pellet loading length was shown for three types of 7 / 8A, 5 / 6A and 1 / 2A. The number of fuel rods is shown as a ratio to 10 rows × 10 columns = 100.

  From this figure, it can be seen that from the viewpoint of securing a pellet loading equivalent to 9 × 9 fuel (type A), if the effective fuel length is 7 / 8A, it can be increased to about 50% (about 50). . In addition, the number of fuel rods can be increased to about 40% (about 40) for 5 / 6A and to about 13% (about 13) for 1 / 2A. Although not shown in this figure, in the case of 2 / 3A, the number of fuel rods can be increased to about 20% (about 20).

  FIG. 2 is a graph showing the relationship between ΔMCPR when the generator load is shut off and the bypass valve is not operating in the end-cycle core of the conventional scrum plant in each embodiment of the fuel assembly according to the present invention, and the number of fuel rods of partially missing fuel rods. It is a graph which shows a relationship. The core 3 is loaded with the same 10 × 10 fuel assembly as in FIG. 3, and the horizontal axis indicates the number of fuel rods when the fuel rods are arranged at all positions of the 10 × 10 square lattice ( 100).

ΔMCPR is 0.26 when the number of fuel rods with a nuclear fuel material loading length shorter than A is zero. When the loaded length of the nuclear fuel material of the partially missing fuel rods 72, 81, 82 is 7 / 8A, even if the number of fuel rods is increased, the effect of mitigating ΔMCPR is hardly obtained. On the other hand, when the loading length of the nuclear fuel material of the partially defective fuel rod is 5 / 6A, ΔMCPR is improved by about 0.05 when the fuel rod is about 20%, and ΔMCPR is about 0.10 when the fuel rod is about 35%. It has been improved. The improvement rate of ΔMCPR is maximized at 1/5 to 1/3 of N ² . Although a greater than 40% of the N ² falls below the feed water heating loss during ΔMCPR (0.15), the improvement rate is small. Further, in some defective fuel loading length of the nuclear fuel material of rods 1 / 2A, improvement of ΔMCPR has a maximum 1 / 10-1 / 5 of ^{N 2.}

In this manner, the loading length of the nuclear fuel material partially deficient fuel rods 71,81,82 is a fuel rod 5 / 6A~1 / 2A a certain percentage loading of ^{N 2,} the pellet loading 9 × 9 The ΔMCPR at the time of transition can be efficiently reduced while being equal to or greater than that of the fuel (type A). Specifically, the nuclear fuel material of the fuel rods of some defective fuel loading length of the nuclear fuel material of rod 5 / 6A~2 / 3A, about 1 / 5-1 / 3 of the N ^2, a part deficient fuel rods About 0 to 1/5 of N ² may be used for a 1 / 2A fuel rod with a loading length of 2 / 3A.

  FIG. 1 is a graph showing the relationship between ΔMCPR at the end of cycle of a conventional scrum plant in the fuel assembly according to the present invention and the volume of a deficient region when the generator load is cut off and the bypass valve is not operating. is there. Here, the volume of the defect regions 91 and 92 is the volume of the nuclear fuel material when the fuel rods loaded with the nuclear fuel material are arranged over the entire range of the effective fuel length at all the positions of the N-by-N square lattice positions. Shown as a percentage of.

  When the loading length of the nuclear fuel material of the partially defective fuel rods 72, 81, 82 is 5 / 6A to 1 / 2A, the volume ratio of the defective region is set to about 3-6%, so that 9 × 9 fuel ΔMCPR can be lowered to 0.20 to 0.15 after securing a pellet loading equal to or greater than (A type). This relationship is also almost true for a 10 × 10 fuel assembly having, for example, two water rods 53 and a water area corresponding to eight fuel rods. Furthermore, the ratio of the number of partially missing fuel rods 72, 81, 82 to the total number of fuel rods of the N × N fuel assembly, the volume of the missing region relative to the pellet loading volume or the cross-sectional area of the N × N assembly, or When viewed as a ratio of the cross-sectional area, it is substantially common to BWR fuels, particularly 9 × 9 fuel assemblies, 10 × 10 fuel assemblies, and 11 × 11 fuel assemblies.

  When the hollow pellet 56 is loaded in the upper region of the fuel assembly 30, the water-to-uranium ratio in the same region increases, so the negative void coefficient decreases and ΔMCPR is reduced. Even when the solid moderator 59 is loaded in the upper region of the fuel assembly 30, the same effect can be obtained. Further, if the outer diameter of the pellet 54 in the upper region of the fuel assembly 30 is reduced, or if the initial helium sealing pressure is reduced, the heat transfer coefficient in the pellet-cladding gap is reduced, so that it is transient. Even when the fuel surface heat flux increases with time when the fuel temperature rises, the time change becomes gentle and the axial output distribution is strained upward, ΔMCPR is reduced.

  When the enrichment average value in the upper region of the fuel assembly 30 is smaller than the enrichment average value in the assembly lower region, or the density average value in the upper region of the fuel assembly 30 is lower in the fuel assembly lower region. When the density average value is smaller, the nuclear reaction at the upper part of the fuel assembly 30 becomes lower than the reaction at the lower part. Therefore, the axial power distribution is less likely to become the upper peak. Further, by making the average gadolinia concentration value in the upper region of the fuel assembly 30 smaller than the average gadolinia concentration value in the lower region of the assembly, the axial output distribution at the end of the cycle may avoid the upper peak. It becomes easy.

  When this method is adopted, the ΔMCPR when the generator load is shut off and the bypass valve is not operated is reduced by a maximum of 0.10 in the end-cycle core of the conventional scrum plant, and the fuel loading volume of the conventional 9 × 9 fuel assembly is reduced. It can be secured.

  Further, the adoption of the partial length fuel rods reduces the fuel assembly pressure loss, thereby improving the hydro-hydraulic stability (channel hydraulic stability). As a result, the stability (core stability, region stability) in which the core and the thermal hydraulic power are combined is improved. At this time, if the partial-length fuel rods are concentrated on the outer periphery of the assembly facing the bypass region, or if they are adjacent to the large-diameter water rod or water channel, the neutron flux is effectively decelerated. Therefore, the negative void coefficient reduction effect is expanded, and the influence of nuclear feedback when the void disappears can be mitigated.

It is a graph which shows the relationship between (DELTA) MCPR at the time of a generator load interruption | blocking and bypass valve non-operation in the cycle end core of the conventional scram plant in each embodiment of the fuel assembly concerning this invention, and the volume of a defect | deletion area | region. The graph which shows the relationship between (DELTA) MCPR at the time of a generator load interruption | blocking and a bypass valve non-operation in the end-cycle core of the conventional scrum plant in each embodiment of the fuel assembly concerning the present invention, and the number of fuel rods of a partly missing fuel rod It is. It is a graph which shows the example of the relationship between the loading amount of a nuclear fuel material, and the number of fuel rods for every loading length of a nuclear fuel material. FIG. 8 is a cross-sectional view taken along arrows Z4-Z4 in FIGS. 6 and 7 schematically showing a cross-section in the first embodiment of the fuel assembly according to the present invention. FIG. 8 is a cross-sectional view taken in the direction of arrows Z5-Z5 in FIGS. 6 and 7 schematically showing a cross-section in the first embodiment of the fuel assembly according to the present invention. FIG. 6 is a vertical cross-sectional view taken along arrows X6-X6 in FIGS. 4 and 5. FIG. 6 is a longitudinal sectional view taken along arrows X7-X7 in FIGS. 4 and 5. It is a graph which shows the example of the core average axial direction power distribution in the cycle initial stage of the core which loaded 1st Embodiment of the fuel assembly based on this invention, and the cycle end stage. It is a graph which shows the transient analysis result at the time of the turbine trip bypass valve operation | movement in the end-cycle core of the conventional scrum plant loaded with 1st Embodiment of the fuel assembly which concerns on this invention. It is a graph which shows the transient analysis result at the time of the turbine trip bypass valve operation | movement in the end-cycle core of the conventional scrum plant loaded with 1st Embodiment of the fuel assembly which concerns on this invention. It is a graph which shows the transient analysis result at the time of the turbine trip bypass valve operation | movement in the end-cycle core of the conventional scrum plant loaded with 1st Embodiment of the fuel assembly which concerns on this invention. It is a graph which shows the core stability reduction ratio of the core loaded with 1st Embodiment of the fuel assembly which concerns on this invention. FIG. 17 is a transverse cross-sectional view taken along arrows Z13-Z13 in FIGS. 15 and 16, schematically showing a cross-section in the second embodiment of the fuel assembly according to the present invention. FIG. 17 is a cross-sectional view taken along arrows Z14-Z14 in FIGS. 15 and 16, schematically showing a cross-section in the second embodiment of the fuel assembly according to the present invention. It is a X15-X15 arrow longitudinal cross-sectional view in FIG. 13 and FIG. It is a X16-X16 arrow longitudinal cross-sectional view in FIG. 13 and FIG. It is a graph which shows the example of the core average axial direction power distribution in the cycle initial stage and the cycle end of the core which loaded the fuel assembly in this Embodiment. FIG. 22 is a cross-sectional view taken along arrows Z18-Z18 in FIGS. 20 and 21, schematically showing a cross-section in the third embodiment of the fuel assembly according to the present invention. FIG. 22 is a transverse cross-sectional view taken along arrows Z19-Z19 in FIGS. 20 and 21, schematically showing a cross-section in the third embodiment of the fuel assembly according to the present invention. FIG. 20 is a vertical cross-sectional view taken along arrow X20-X20 in FIGS. 18 and 19. It is a X21-X21 arrow longitudinal cross-sectional view in FIG.18 and FIG.19. FIG. 26 is a cross-sectional view taken along arrows Z22-Z22 in FIGS. 24 and 25, schematically showing a cross-section in the fourth embodiment of the fuel assembly according to the present invention. FIG. 26 is a cross-sectional view taken along arrows Z23-Z23 in FIGS. 24 and 25, schematically showing a cross-section in the fourth embodiment of the fuel assembly according to the present invention. It is a X24-X24 arrow longitudinal cross-sectional view in FIG. 22 and FIG. It is a X24-X24 arrow longitudinal cross-sectional view in FIG. 22 and FIG. FIG. 26 is a cross-sectional view taken along arrows Z26-Z26 in FIGS. 24 and 25, schematically showing a cross-section in the fifth embodiment of the fuel assembly according to the present invention. It is a Z27-Z27 arrow transverse cross section in Drawing 24 and Drawing 25 showing typically a cross section in a 5th embodiment of a fuel assembly concerning the present invention. It is a X28-X28 arrow longitudinal cross-sectional view in FIG.22 and FIG.23. It is a X29-X29 arrow longitudinal cross-sectional view in FIG.22 and FIG.23. It is a Z30-Z30 arrow transverse cross section in Drawing 32 and Drawing 33 showing typically a cross section in a 6th embodiment of a fuel assembly concerning the present invention. FIG. 34 is a transverse sectional view taken along arrows Z31-Z31 in FIGS. 32 and 33, schematically showing a transverse section in the sixth embodiment of the fuel assembly according to the present invention. FIG. 32 is a longitudinal sectional view taken along arrows X32-X32 in FIGS. 30 and 31. FIG. 32 is a longitudinal sectional view taken along arrows X33-X33 in FIGS. 30 and 31. It is a graph which shows the analysis result of MCPR of the core loaded with 6th Embodiment of the fuel assembly which concerns on this invention. It is a Z35-Z35 arrow transverse cross section in Drawing 37 and Drawing 38 showing typically a cross section in a 7th embodiment of a fuel assembly concerning the present invention. It is a Z36-Z36 arrow cross-sectional view in Drawing 37 and Drawing 38 showing typically a cross section in a 7th embodiment of a fuel assembly concerning the present invention. It is a X37-X37 arrow longitudinal cross-sectional view in FIG. 35 and FIG. It is a X38-X38 arrow longitudinal cross-sectional view in FIG. 35 and FIG. It is a partial expanded longitudinal cross-sectional view of the segment in 7th Embodiment of the fuel assembly which concerns on this invention. It is a Z40-Z40 arrow transverse cross section in Drawing 42 and Drawing 43 showing typically a cross section in an 8th embodiment of a fuel assembly concerning the present invention. FIG. 44 is a cross-sectional view taken along arrows Z41-Z41 in FIGS. 42 and 43, schematically showing a cross-section in the eighth embodiment of the fuel assembly according to the present invention. It is a X42-X42 arrow longitudinal cross-sectional view in FIG. 40 and FIG. It is a X43-X43 arrow longitudinal cross-sectional view in FIG. 40 and FIG. It is a Z44-Z44 arrow transverse cross section in Drawing 46 and Drawing 47 showing typically a cross section in an 8th embodiment of a fuel assembly concerning the present invention. It is a Z45-Z45 arrow transverse cross section in Drawing 46 and Drawing 47 showing typically a cross section in an 8th embodiment of a fuel assembly concerning the present invention. It is a X46-X46 arrow longitudinal cross-sectional view in FIG.44 and FIG.45. It is a X47-X47 arrow longitudinal cross-sectional view in FIG.44 and FIG.45. FIG. 52 is a transverse sectional view taken along arrows Z48-Z48 in FIGS. 50 and 51, schematically showing a transverse section in the eighth embodiment of the fuel assembly according to the present invention. FIG. 52 is a transverse cross-sectional view taken along arrows Z49-Z49 in FIGS. 50 and 51, schematically showing a transverse section in the eighth embodiment of the fuel assembly according to the present invention. It is a X50-X50 arrow longitudinal cross-sectional view in FIG. 48 and FIG. It is a X51-X51 arrow longitudinal cross-sectional view in FIG. 48 and FIG. It is a block diagram shown with the typical cross section of a part of typical boiling water nuclear power plant (BWR). It is a figure which shows the typical driving | running characteristic of BWR. It is a cross-sectional view schematically showing a typical BWR core. 1 is a longitudinal sectional view schematically showing a typical BWR core. FIG. It is a partially cutaway perspective view of a 9 × 9 fuel assembly (A type). It is a partially expanded partly cutaway perspective view of a 9 × 9 fuel assembly (type A). It is a cross-sectional view which shows a 9x9 fuel assembly (A type) with a control rod. It is a graph which shows the example of the core average axial direction power distribution in the cycle initial stage and the cycle end about the core loaded with the 9x9 fuel assembly (A type). It is a graph which shows the turbine trip bypass valve malfunction analysis result regarding the BWR plant using the conventional scrum loaded with all 9 × 9 fuel A types. It is a graph which shows the turbine trip bypass valve malfunction analysis result regarding the BWR plant using the conventional scrum loaded with all 9 × 9 fuel A types. It is a graph which shows the turbine trip bypass valve malfunction analysis result regarding the BWR plant using the conventional scrum loaded with all 9 × 9 fuel A types. It is a graph which shows the core stability analysis result of the BWR plant which loaded all the 9x9 fuel (A type). It is a partially cutaway perspective view of a 9 × 9 fuel assembly (B type). It is a cross-sectional view which shows a 9x9 fuel assembly (B type) with a control rod. It is the table | surface which put together the characteristic of each embodiment.

Explanation of symbols

DESCRIPTION OF SYMBOLS 1 ... Reactor containment vessel, 2 ... Reactor pressure vessel, 3 ... Core, 4 ... Cooling water, 8 ... Recirculation system, 9 ... Main steam piping system, 10 ... Turbine, 16 ... Condenser, 17 ... Condensate Pump, 18 ... Feed heater, 19 ... Feed pump, 30 ... Fuel assembly, 31 ... Channel box, 32 ... Control rod, 33 ... LPRM, 34 ... Core support plate, 35 ... Upper grid plate, 36 ... Shroud, 37 ... Fuel support bracket, 51 ... Fuel rod, 52 ... Water channel, 53 ... Water rod, 54 ... Pellet, 55 ... Clad tube, 56 ... Hollow pellet, 57 ... Small diameter pellet, 58 ... Low concentrated pellet, 61 ... Upper type 62, lower tie plate, 63 ... spacer, 71 ... full length fuel rod, 72 ... full length partially missing fuel rod, 73 ... upper hollow fuel rod, 74 ... upper narrow fuel rod, 75 ... upper low enrichment fuel rod, 76 ... Upper part Fuel rod with body moderator, 81, 82 ... Short partially missing fuel rod (partial length fuel rod), 83, 84 ... Segment, 85 ... Segment joint, 86 ... Plenum spring, 91, 92 ... Lost region, 93 ... Plenum

Claims (24)

In the fuel assembly loaded in the boiling water reactor with nuclear fuel material contained in the axially effective fuel effective part,
A fuel rod in which the nuclear fuel material is housed in at least a part of the fuel effective portion of a cylindrical cladding tube arranged in at least a part of a square lattice position of N rows and N columns in a direction in which a plane perpendicular to the axial direction extends. When,
A lower tie plate that supports the fuel rod at the lower end;
An upper tie plate that supports at least a portion of the fuel rod at its upper end;
The fuel rod includes a full length fuel rod that contains the nuclear fuel material over the entire fuel effective portion, and a deficient region that does not contain the nuclear fuel material extends downward from an upper end of the fuel effective portion. A partially deficient fuel rod formed in a range extending 1/6 or more of the fuel effective length which is the length of the fuel effective portion, and the full length fuel rod is arranged at all positions of the square lattice position The fuel assembly, wherein a ratio of the volume of the defect region to the volume of the nuclear fuel material is 3% or more and 6% or less. In the fuel assembly loaded in the boiling water reactor with nuclear fuel material contained in the axially effective fuel effective part,
A fuel rod in which the nuclear fuel material is housed in at least a part of the fuel effective portion of a cylindrical cladding tube arranged in at least a part of a square lattice position of N rows and N columns in a direction in which a plane perpendicular to the axial direction extends. When,
A lower tie plate that supports the fuel rod at the lower end;
An upper tie plate that supports at least a portion of the fuel rod at its upper end;
The fuel rod includes a full length fuel rod that contains the nuclear fuel material over the entire fuel effective portion, and a deficient region that does not contain the nuclear fuel material extends downward from an upper end of the fuel effective portion. A number of partially deficient fuel rods in the range of 1/5 to 1/3 of N ² formed in a range extending 1/6 or more of the fuel effective length which is the length of the fuel effective portion, Fuel assembly. The partially deficient fuel rod includes a first partially deficient fuel rod formed in a range in which the deficient region extends from 1/6 to 1/3 below the effective fuel length from the upper end of the effective fuel portion; The number of second partially deficient fuel rods equal to or less than 1/5 of N ² formed in a range in which the deficient region extends from 1/3 to 1/2 of the effective fuel length downward from the upper end of the effective fuel portion. The fuel assembly according to claim 2, comprising:   The fuel assembly according to any one of claims 1 to 3, wherein the partially missing fuel rod includes a short partially missing fuel rod whose upper end is located apart from the upper tie plate.   A plenum is formed in the vicinity of the upper ends of the full length fuel rod and the short partly missing fuel rod, and the plenum length of the short partly missing fuel rod is equal to or less than the plenum length of the full length fuel rod. The fuel assembly according to claim 4.   The partially missing fuel rod includes a full length partially missing fuel rod whose upper end is supported by the upper tie plate, and at least one of the short partially missing fuel rods has the same axial position as the full length partially missing fuel rod. The fuel assembly according to claim 4, wherein the defect region is formed in the fuel assembly.   The fuel assembly according to any one of claims 1 to 6, wherein a solid moderator is accommodated in the defect region. In the fuel assembly loaded in the boiling water reactor with nuclear fuel material contained in the axially effective fuel part,
A fuel rod in which the nuclear fuel material is housed in at least a part of the fuel effective portion of a cylindrical cladding tube arranged in at least a part of a square lattice position of N rows and N columns in a direction in which a plane perpendicular to the axial direction extends. When,
A lower tie plate that supports the fuel rod at its lower end;
An upper tie plate that supports at least a portion of the fuel rod at its upper end;
The fuel rods contain the nuclear fuel material over the whole fuel effective portion, and at least 1/12 above the fuel effective length which is the length of the fuel effective portion and the lower end of the fuel effective portion. A full-length fuel rod containing the nuclear fuel material other than uranium having a concentration of natural uranium or less in a range from a distant position to a position distant from the upper end of the effective fuel portion and 1/12 of the effective fuel length; A deficient region containing uranium with a concentration of natural uranium or less is formed in a range extending from the upper end of the fuel effective portion downward to 1/6 or more of the fuel effective length and is 1/5 or more and 1/3 or less of N ² The ratio of the volume of the deficient region to the volume of the nuclear fuel material when the full length fuel rods are arranged at all positions of the square lattice position is 3% or more and 6% or less. A fuel assembly characterized by body. In the fuel assembly loaded in the boiling water reactor with nuclear fuel material contained in the axially effective fuel effective part,
A fuel rod in which the nuclear fuel material is housed in at least a part of the fuel effective portion of a cylindrical cladding tube arranged in at least a part of a square lattice position of N rows and N columns in a direction in which a plane perpendicular to the axial direction extends. When,
A lower tie plate that supports the fuel rod at the lower end;
An upper tie plate that supports at least a portion of the fuel rod at its upper end;
The fuel rods contain the nuclear fuel material over the whole fuel effective portion, and at least 1/12 above the fuel effective length which is the length of the fuel effective portion and the lower end of the fuel effective portion. A full-length fuel rod containing the nuclear fuel material other than uranium having a concentration of natural uranium or less in a range from a distant position to a position distant from the upper end of the effective fuel portion and 1/12 of the effective fuel length; A deficient region containing uranium with a concentration of natural uranium or less is formed in a range extending from the upper end of the fuel effective portion downward to 1/6 or more of the fuel effective length and is 1/5 or more and 1/3 or less of N ² A fuel assembly comprising a number of partially deficient fuel rods. In the fuel assembly loaded in the boiling water reactor with nuclear fuel material contained in the axially effective fuel part,
A fuel rod in which the nuclear fuel material is housed in at least a part of the fuel effective portion of a cylindrical cladding tube arranged in at least a part of a square lattice position of N rows and N columns in a direction in which a plane perpendicular to the axial direction extends. When,
A lower tie plate that supports the fuel rod at its lower end;
An upper tie plate that supports at least a portion of the fuel rod at its upper end;
The fuel rod includes a full length fuel rod containing solid pellets of the nuclear fuel material over the entire fuel effective portion, and a length of the fuel effective portion upward from a lower end of the fuel effective portion. The solid pellets are housed in a first range extending from ½ to 5/6 of an effective fuel length, and the solid pellet is in a second range from the upper end of the first range to the upper end of the effective fuel portion. An upper hollow fuel rod having a number equal to or greater than 1/3 of N ² and containing a hollow pellet having the same outer diameter as the pellet and having an inner diameter equal to or smaller than ¼ of the outer diameter of the solid pellet. A fuel assembly.   2. The fuel rod according to claim 1, wherein the fuel rod includes a short partially-defective fuel rod in which the solid pellet is accommodated in the same axial range as the first range and an upper end is located apart from the upper tie plate. Item 9. The fuel assembly according to Item 8. In the fuel assembly loaded in the boiling water reactor with nuclear fuel material contained in the axially effective fuel effective part,
A fuel rod in which the nuclear fuel material is housed in at least a part of the fuel effective portion of a cylindrical cladding tube arranged in at least a part of a square lattice position of N rows and N columns in a direction in which a plane perpendicular to the axial direction extends. When,
A lower tie plate that supports the fuel rod at the lower end;
An upper tie plate that supports at least a portion of the fuel rod at its upper end;
The fuel rod includes a full length fuel rod containing the first solid pellet of the nuclear fuel material over the whole fuel effective portion, and the fuel effective portion above the lower end of the fuel effective portion. The first solid pellet is accommodated in a first range extending from ½ to 5/6 of the effective fuel length which is a length, and is in a range from the upper end of the first range to the upper end of the effective fuel portion. An upper small-diameter fuel rod having a number equal to or more than 1/3 of N ² and containing a ^second solid pellet having an outer diameter of 90% or less of the outer diameter of the first solid pellet. Characteristic fuel assembly.   The fuel rod includes a short partially missing fuel rod in which the first solid pellet is accommodated in the same axial range as the first range and an upper end is located apart from the upper tie plate. The fuel assembly according to claim 12. In the fuel assembly loaded in the boiling water reactor with nuclear fuel material contained in the axially effective fuel effective part,
A fuel rod in which the nuclear fuel material is housed in at least a part of the fuel effective portion of a cylindrical cladding tube arranged in at least a part of a square lattice position of N rows and N columns in a direction in which a plane perpendicular to the axial direction extends. When,
A lower tie plate that supports the fuel rod at the lower end;
An upper tie plate that supports at least a portion of the fuel rod at its upper end;
And the fuel rods contain helium in a range extending from ½ to 5/6 of the effective fuel length, which is the length of the effective fuel portion, upward from the lower end of the effective fuel portion. The first fuel segment is connected to the upper end of the first segment, and the nuclear fuel material is accommodated in a range up to the upper end of the fuel effective portion, so that the initial helium sealing pressure of the first segment is 50% or less. A fuel assembly, comprising: an upper low-pressure fuel rod having a number equal to or more than 1/3 of N ² encapsulating helium with pressure.   The fuel rod contains the nuclear fuel material downward from the same position in the axial direction as the coupling position of the first segment and the second segment, and helium at a pressure equal to the initial helium filling pressure of the first segment. The fuel assembly according to claim 14, wherein the fuel assembly includes a short partly deficient fuel rod whose upper end enclosing a gas is positioned away from the upper tie plate. In the fuel assembly loaded in the boiling water reactor with nuclear fuel material contained in the axially effective fuel part,
A fuel rod in which the nuclear fuel material is housed in at least a part of the fuel effective portion of a cylindrical cladding tube arranged in at least a part of a square lattice position of N rows and N columns in a direction in which a plane perpendicular to the axial direction extends. When,
A lower tie plate that supports the fuel rod at the lower end;
An upper tie plate that supports at least a portion of the fuel rod at its upper end;
The fuel rods contain the nuclear fuel material in a first range extending from ½ to 5/6 of the effective fuel length that is the length of the effective fuel portion upward from the lower end of the effective fuel portion. N containing the nuclear fuel material in a second range from the upper end of the range to the upper end of the effective fuel portion so that the average enrichment is 1 / 1.1 or less of the average enrichment of the first range. ^And a number of upper low-enriched fuel rods having a number of 1/3 or more. In the fuel assembly loaded in the boiling water reactor with nuclear fuel material contained in the axially effective fuel effective part,
A fuel rod in which the nuclear fuel material is housed in at least a part of the fuel effective portion of a cylindrical cladding tube arranged in at least a part of a square lattice position of N rows and N columns in a direction in which a plane perpendicular to the axial direction extends. When,
A lower tie plate that supports the fuel rod at the lower end;
An upper tie plate that supports at least a portion of the fuel rod at its upper end;
The fuel rod has a first range of the nuclear fuel material in a first range extending from ½ to 5/6 of the fuel effective length that is the length of the fuel effective portion upward from the lower end of the fuel effective portion. A second range of the nuclear fuel material having an average density of 1 / 1.1 or less of the average density of the first pellets in a second range from the upper end of the range to the upper end of the effective fuel portion. A fuel assembly comprising: an upper low-density fuel rod having a number equal to or more than 1/3 of N ² containing pellets. In the fuel assembly loaded in the boiling water reactor with nuclear fuel material contained in the axially effective fuel effective part,
A fuel rod in which the nuclear fuel material is housed in at least a part of the fuel effective portion of a cylindrical cladding tube arranged in at least a part of a square lattice position of N rows and N columns in a direction in which a plane perpendicular to the axial direction extends. When,
A lower tie plate that supports the fuel rod at the lower end;
An upper tie plate that supports at least a portion of the fuel rod at its upper end;
The fuel rods contain the nuclear fuel material in a first range extending from ½ to 5/6 of the effective fuel length that is the length of the effective fuel portion upward from the lower end of the effective fuel portion. The nuclear fuel material and gadolinia are stored in a second range from the upper end of the range to the upper end of the effective fuel portion so that the average gadolinia concentration is 1 / 1.1 or less of the average gadolinia concentration of the first range. And N ² or more upper low-enriched fuel rods. The water rod is disposed near the center of the square lattice position and has a maximum cross-sectional area of 7 to 10 times the cross-sectional area of the fuel rod, the N is 9 to 11 and the nuclear fuel material is Pellets with a maximum density of 10.6 g / cm ³ or less, the maximum enrichment of the nuclear fuel material is less than 5%, and the fuel rods are filled with helium at an initial helium fill pressure of 10 atmospheres or less, The fuel assembly according to any one of claims 1 to 18, wherein the effective fuel length is 3.6 m or more and less than 3.9 m.   N is 10, and when each of the square lattice positions is expressed as (i, j) using row number i and column number j, (5, 5), (5, 6), ( 5,7), (6,5), (6,6), (6,7), (7,5), (7,6) and water arranged at the position represented by (7,7) The fuel rods contain the nuclear fuel material in a range extending about 3.0 m upward from the lower end of the fuel effective part (1, 4), (1, 5), (1, 6), (1,7), (4,1), (5,1), (6,1), (7,1), (10,4), (10,5), (10,6), (10 , 7), (4, 10), (5, 10), (6, 10), (7, 10), (4, 5), (4, 7), (5, 4), (5, 7 ), (7,5) and (7,7) Including location is 24 parts active length fuel rods have, fuel assembly according to any one of claims 1 to 19, characterized in that.   N is 10, and when each of the square lattice positions is expressed as (i, j) using row number i and column number j, (5, 5), (5, 6), ( 5,7), (6,5), (6,6), (6,7), (7,5), (7,6) and water arranged at the position represented by (7,7) The fuel rods contain the nuclear fuel material in a range extending about 3.0 m upward from the lower end of the fuel effective part (1, 4), (1, 5), (1, 6), (1,7), (4,10), (5,10), (6,10), (4,1), (5,1), (6,1), (7,1), (10 , 4), 14 first partially effective length fuel rods disposed at positions indicated by (10, 5) and (10, 6), respectively, and about 2 upward from the lower end of the fuel effective portion. In the range extending 0.0 m The fuel material is contained and disposed at each of the positions represented by (4,5), (4,7), (5,8), (5,4), (7,4) and (8,5). The fuel assembly according to any one of claims 1 to 19, further comprising six second partial effective length fuel rods.   N is 10, and when each of the square lattice positions is represented as (i, j) using a row number i and a column number j, (4, 4), (4, 5), ( 4,6), (5,4), (5,5), (5,6), (6,4), (6,5) and water arranged at the position represented by (6,6) The fuel rods contain the nuclear fuel material in a range extending about 3.0 m upward from the lower end of the effective fuel portion (1, 5), (1, 6), (1, 7), (4,10), (5,10), (6,10), (7,10), (5,1), (6,1), (7,1), (10,4), (10 , 5), 14 first partially effective length fuel rods arranged at the positions represented by (10, 6) and (10, 7), respectively, and about 2 upward from the lower end of the effective fuel portion. In the range that extends 0.0m Contains nuclear fuel material and is placed at each of the positions represented by (3,6), (4,7), (6,7), (6,3), (7,4) and (7,6) The fuel assembly according to any one of claims 1 to 19, comprising six second partial effective length fuel rods.   N is 10, and when each of the square lattice positions is represented as (i, j) using row number i and column number j, (4, 6), (4, 7), ( 5, 6) and (5, 7) and the positions (6, 4), (6, 5), (7, 4) and (7, 5). The fuel rods contain the nuclear fuel material in a range extending about 3.0 m upward from the lower end of the fuel effective portion (1, 4), (1, 5), (1,6), (1,7), (4,10), (5,10), (6,10), (7,10), (4,1), (5,1), (6 , 1), (7, 1), (10, 4), (10, 5), (10, 6) and 16 first arranged at each of the positions represented by (10, 7) A partial effective length fuel rod and the fuel (4,5), (5,4), (5,5), (6,6), (6,7) and the nuclear fuel material in a range extending about 2.0 m upward from the lower end of the effective portion 20. The six second partial effective length fuel rods disposed at each of the positions represented by (7, 7), respectively, according to any one of claims 1 to 19. The fuel assembly as described. The N is 10, and the nuclear fuel material of about 1.90 × 10 ⁴ cm ³ or more which is the volume of the nuclear fuel material contained in 9 × 9 fuel (A type) is contained in the fuel rod. The fuel assembly according to any one of claims 1 to 23, wherein the fuel assembly is any one of claims 1 to 23.

Images