阅读说明：本技术 一种循环长度灵活调节的压水堆堆芯燃料管理方法 (Method for managing fuel in pressurized water reactor core with flexibly adjusted cycle length ) 是由 魏盛辉 徐琳琳 高海滨 张瑜 王释伟 张伟斌 于 2021-09-17 设计创作，主要内容包括：本发明涉及一种循环长度灵活调节的压水堆堆芯燃料管理方法,对于燃料组件总数不变的压水堆堆芯,原本采用固定新燃料组件数为N-(0)组的换料策略,根据目标循环长度的延长或缩短需求,相应调整下一轮循环的新燃料组件数为N-(0)+4n或N-(0)－4n；其中,n为1或2。本发明所述新型压水堆堆芯燃料管理方法,可以对循环长度进行灵活调节,解决了因循环长度相对固定而导致的调节停堆时间灵活性较差、经济性较差等问题,例如可以将原来固定的换料周期18个月调整为17～19个月,甚至16～20个月,而且不影响技术规格书中各项定期试验的执行。(The invention relates to a method for managing fuel in a pressurized water reactor core with flexibly adjusted cycle length, which is characterized in that the number of originally adopted fixed new fuel assemblies is N for the pressurized water reactor core with invariable total number of fuel assemblies 0 The group refueling strategy correspondingly adjusts the number of new fuel assemblies of the next cycle to be N according to the requirement of prolonging or shortening the target cycle length 0 +4N or N 0 -4 n; wherein n is 1 or 2. The novel method for managing the fuel in the reactor core of the pressurized water reactor can flexibly adjust the circulating length, solves the problems of poor flexibility of adjusting the shutdown time, poor economy and the like caused by relatively fixed circulating length, can adjust the original fixed refueling period of 18 months to 17-19 months, even 16-20 months, and does not influence the execution of each periodic test in the technical specification.)

1. A method for managing fuel in a pressurized water reactor core with flexibly adjusted cycle length is characterized in that for the pressurized water reactor core with invariable total number of fuel assemblies, the number of originally adopted fixed new fuel assemblies is N₀The group refueling strategy correspondingly adjusts the number of new fuel assemblies of the next cycle to be N according to the requirement of prolonging or shortening the target cycle length₀+4N groups or N₀-4 n groups; wherein n is 1 or 2.

2. The method for managing the fuel in the pressurized water reactor core with the flexibly adjusted cycle length as claimed in claim 1, wherein the enrichment degree of a new fuel assembly of a next cycle is calculated according to the demand of the flattening power distribution and based on the initial enrichment degree, the fuel consumption and the residual enrichment degree of the fuel of the current cycle in operation and the demand of the energy demand and reactivity control of the next cycle, and the enrichment degree of the new fuel assembly is less than 5%.

3. The method for managing the fuel in the core of the pressurized water reactor with the flexibly adjusted cycle length as claimed in claim 2, wherein the new fuel assemblies in the next cycle are calculated according to the demand of the flattening power distribution and based on the initial enrichment, the burnup and the residual enrichment of the fuel in the current cycle in operation and the requirements of the energy demand and reactivity control of the next cycle, wherein the two new fuel assemblies are different in enrichment degree and are respectively a% and b%, and a% < b% < 5%.

4. The method for managing fuel in a pressurized water reactor core with flexibly adjusted cycle length according to claim 3, wherein the number of new fuel assemblies with the enrichment degree of a% is N₁The number of the groups is set to be,the number of new fuel assemblies with the enrichment degree of b% is N₂Group (b) wherein N₁≥N₂And N is₁And N₂Are all multiples of 4, N₁And N₂Is a sum of N₀+4N or N₀－4n。

5. The method for managing the fuel in the pressurized water reactor core with the flexibly adjusted cycle length according to any one of claims 1 to 4, characterized in that for the pressurized water reactor core with the total number of fuel assemblies of 157 groups, a refueling strategy with a fixed number of new fuel assemblies of 64 groups is originally adopted, and the number of new fuel assemblies in the next cycle is correspondingly adjusted to be 64+4n groups or 64-4 n groups according to the requirement of prolonging or shortening the target cycle length; wherein N is₀N is 1 or 2, 64.

6. The method for managing the fuel in the core of the pressurized water reactor with the flexibly adjusted cycle length according to the claim 5, is characterized in that the new fuel assemblies of the next cycle are calculated to be two new fuel assemblies with the enrichment degrees of 4.45% and 4.95% respectively according to the requirement of the flattening power distribution and based on the initial enrichment degree, the fuel consumption and the residual enrichment degree of the fuel of the current cycle in operation and the requirements of the energy requirement and the reactivity control of the next cycle, namely, a is 4.45, and b is 4.95.

7. The method for managing the fuel in the pressurized water reactor core with the flexibly adjusted cycle length as recited in claim 6, wherein for the number of new fuel assemblies in the next cycle of 64+4n groups or 64-4 n groups, the number of new fuel assemblies with the enrichment degree of 4.45% is 36 groups, and the number of new fuel assemblies with the enrichment degree of 4.95% is 28+4n groups or 28-4 n groups; wherein N is₁＝36，N₂28+4n or 28-4 n, n is 1 or 2.

8. The method for fuel management of a pressurized water reactor core with flexibly adjusted cycle length according to any one of claims 3 to 7, wherein the next cycle adopts a low leakage loading mode, old fuel assemblies from a previous cycle are arranged at the outermost circle, new fuel assemblies with b% enrichment are arranged at the next outer circle, and new fuel assemblies with a% enrichment and the rest old fuel assemblies from the previous cycle are arranged in the inner region in a checkerboard manner.

9. The method for flexibly adjusting the fuel of the pressurized water reactor core according to any one of claims 1 to 8, wherein the model of the pressurized water reactor core is CAP1000 or AP 1000.

10. The method for managing fuel in a pressurized water reactor core with flexibly adjusted cycle length according to any one of claims 1 to 9, wherein the pressurized water reactor core adopts an integral fuel burnable poison IFBA, and a layer of ZrB is coated on the surface of fuel pellets₂As a flammable absorber.

Technical Field

The invention relates to the field of nuclear power plant reactor core design specialties, in particular to the technical field of pressurized water reactor core fuel management, and particularly relates to a method for managing the pressurized water reactor core fuel with flexibly adjusted cycle length.

Background

A commercial pressurized water reactor core at home and abroad can load a plurality of groups of nuclear fuel assemblies, a first cycle generally loads all new fuel assemblies, the nuclear fuel assemblies continuously fission and release energy, and after a certain time (generally 12-24 months), the reactor core is difficult to continuously maintain fission due to insufficient total reactivity of the reactor core. At this time, control rods are generally inserted to keep the reactor core in a deep subcritical state, the temperature and pressure of a primary circuit are reduced, an upper cover of a pressure vessel is opened, a plurality of groups of fuel assemblies with low reactivity and deep burnup are removed, a plurality of groups of new fuel assemblies with high reactivity are loaded, the new fuel assemblies and the old fuel assemblies are loaded and arranged, and the reactor can be restarted after the completion of refueling overhaul (about 30 days).

The pressurized water reactor nuclear power plant must be shut down and refueled at the end of each cycle life. In order to ensure the cycle length, economy and safety of the next cycle, the refueling core design and safety analysis are completed before each cycle is started. The length of the cycle length needs to consider many factors such as regional energy supply characteristics, power generation requirements, fuel economy, electricity price, realization capability and safety of equipment and systems, convenience in overhaul and the like. At present, a refueling strategy for fixing the number of new fuel assemblies is mostly adopted in a pressurized water reactor nuclear power plant, so that the cycle length of each cycle is relatively fixed, and particularly the cycle length of most of the current pressurized water reactor cores in China is 18 months. However, based on the demand of the power grid and the special production planning of the nuclear power plant, such as heating in winter, etc., major repair cannot be arranged in summer and winter every year, and unexpected factors caused by the power grid or the unit during the operation of the unit require the unit to be changed in advance or postponed, if a mode of changing the material in 18 months with a constant cycle length is adopted, the unit is inevitably arranged to be maintained in summer or spring festival within a few years, and unexpected property caused by the power grid or the unit cannot be coped with.

Therefore, a new method for managing fuel in a pressurized water reactor core is needed, and the circulation length can be flexibly adjusted.

Disclosure of Invention

In view of the problems in the prior art, the invention provides a method for managing fuel in a pressurized water reactor core with flexibly adjusted cycle length, which realizes flexibly-circulated refueling by increasing or reducing the number of new fuel assemblies required by 4n (n is 1 or 2) groups of refueling, can flexibly adjust the cycle length, and solves the problems of poor flexibility of adjusting shutdown time, poor economy and the like caused by relatively fixed cycle length.

In order to achieve the purpose, the invention adopts the following technical scheme:

the invention aims to provide a method for managing fuel in a pressurized water reactor core with flexibly adjusted cycle length, which is used for the pressurized water reactor core with invariable total number of fuel assemblies and adopts N number of fixed new fuel assemblies₀The group refueling strategy correspondingly adjusts the number of new fuel assemblies of the next cycle to be N according to the requirement of prolonging or shortening the target cycle length₀+4N groups or N₀-4 n groups; wherein n is 1 or 2.

The invention provides a novel method for managing fuel in a reactor core of a pressurized water reactor, which can flexibly adjust the circulating length, solves the problems of poor flexibility of adjusting shutdown time, poor economy and the like caused by relatively fixed circulating length, can adjust the original fixed refueling period of 18 months to 17-19 months, even 16-20 months, and does not influence the execution of each periodic test in a technical specification.

It should be noted that, the method for managing fuel in a pressurized water reactor core with flexibly adjusted cycle length according to the present invention is directed to a pressurized water reactor core with a constant total number of fuel assemblies, such as 121 groups or 157 groups, and when the number of new fuel assemblies in a next cycle is increased or decreased by 4n, the number of old fuel assemblies from a previous cycle is correspondingly decreased or increased by 4n, so as to ensure that the total number of fuel assemblies in the pressurized water reactor core is constant.

As a preferred technical scheme of the invention, according to the requirement of flattening power distribution, based on the initial enrichment degree, the fuel consumption and the residual enrichment degree of the fuel of the current running cycle and the requirement of the energy requirement and reactivity control of the next cycle, the enrichment degree of the new fuel assembly of the next cycle is calculated, and the enrichment degree of the new fuel assembly is less than 5 percent, because the enrichment degree of the new fuel assembly needs to meet the regulatory requirement of 5 percent upper limit of the enrichment degree of the fuel of the civil nuclear facility.

As a preferred technical scheme of the invention, according to the requirement of flattening power distribution, based on the initial enrichment degree, the fuel consumption and the residual enrichment degree of the fuel of the current cycle in operation, and the energy requirement and the reactivity control requirement of the next cycle, the new fuel assemblies of the next cycle are calculated to be two new fuel assemblies with different enrichment degrees and respectively a% and b%, wherein a% is less than b% and less than 5%.

It is worth mentioning that the value of b is as close to 5 as possible, for example, b is 4.95, which can improve the fuel consumption of the fuel assembly, and thus the economy of the nuclear power plant.

As a preferred technical scheme of the invention, the number of the new fuel components with the enrichment degree of a% is N₁The number of new fuel assemblies with b% enrichment is N₂Group (b) wherein N₁≥N₂And N is₁And N₂Are all multiples of 4, N₁And N₂Is a sum of N₀+4N or N₀－4n。

As a preferred technical scheme of the invention, for a pressurized water reactor core with 157 groups of total fuel assemblies, a refueling strategy with 64 groups of fixed new fuel assemblies is originally adopted, and the number of the new fuel assemblies in the next cycle is correspondingly adjusted to be 64+4n groups or 64-4 n groups according to the requirement of prolonging or shortening the target cycle length; wherein N is₀N is 1 or 2, 64.

According to the demand of the flattening power distribution, based on the initial fuel enrichment degree, the fuel consumption and the residual enrichment degree of the running current cycle and the requirements of the next cycle energy demand and reactivity control, the new fuel assemblies of the next cycle are calculated to be two new fuel assemblies with the enrichment degrees of 4.45% and 4.95%, namely, a is 4.45% and b is 4.95%.

As a preferred technical scheme of the invention, the number of new fuel assemblies in the next cycle is 64+4n groups or 64-4 n groups, the number of new fuel assemblies with the enrichment degree of 4.45 percent is 36 groups, and the enrichment is carried outThe number of new fuel assemblies with the degree of 4.95 percent is 28+4n groups or 28-4 n groups; wherein N is₁＝36，N₂28+4n or 28-4 n, n is 1 or 2.

In a preferred embodiment of the present invention, the next cycle uses a low leakage loading mode, wherein old fuel assemblies from the previous cycle are arranged in the outermost circle, new fuel assemblies with b% enrichment are arranged in the next outer circle, and new fuel assemblies with a% enrichment and the remaining old fuel assemblies from the previous cycle are arranged in the inner area in a checkerboard manner.

As a preferable technical scheme of the invention, the model of the pressurized water reactor core is CAP1000 type or AP1000 type, wherein the CAP1000 type is AP1000 type localization, and the design of the core is the same as that of the AP1000 type.

As the preferable technical scheme of the invention, the pressurized water reactor core adopts integral fuel burnable poison IFBA, and a layer of ZrB is coated on the surface of a fuel pellet₂As a flammable absorber.

Compared with the prior art, the invention at least has the following beneficial effects:

the invention provides a novel method for managing fuel in a reactor core of a pressurized water reactor, which realizes flexible circulation refueling by increasing or reducing the number of new fuel assemblies required by 4n (n is 1 or 2) groups of refueling, can flexibly adjust the circulation length, solves the problems of poor flexibility of reactor shutdown time adjustment, poor economy and the like caused by relatively fixed circulation length, can adjust the original fixed refueling period of 18 months to 17-19 months, even 16-20 months, and does not influence the execution of each periodic test in a technical specification.

Drawings

FIG. 1 is a loading scheme of a second cycle core of a Haiyang nuclear power plant according to example 1 of the present invention;

FIG. 2 is a schematic diagram of a flexible cycle fuel management strategy for Haiyang nuclear power plant according to example 1 of the present invention using 4 new fuel assemblies that are continuously and alternately added and subtracted;

FIG. 3 is a loading scheme of a third cycle core of the Haiyang nuclear power plant according to example 1 of the present invention;

FIG. 4 is a loading scheme of a core of a fourth cycle of the Haiyang nuclear power plant according to example 1 of the present invention;

FIG. 5 shows a fifth cycle core loading scheme for the Haiyang nuclear power plant according to example 1 of the present invention;

FIG. 6 is a loading scheme of a core of a sixth cycle of the Haiyang nuclear power plant according to example 1 of the present invention;

FIG. 7 shows a core loading scheme of a seventh cycle of the Haiyang nuclear power plant according to example 1 of the present invention;

FIG. 8 shows a loading scheme of a core of an eighth cycle of the Haiyang nuclear power plant according to embodiment 1 of the present invention.

Detailed Description

The technical scheme of the invention is further explained by the specific implementation mode in combination with the attached drawings.

To better illustrate the invention and to facilitate the understanding of the technical solutions thereof, typical but non-limiting examples of the invention are as follows:

example 1

The embodiment provides a pressurized water reactor core fuel management strategy with flexibly adjusted cycle length, which takes a Haiyang nuclear power plant as a research object and specifically comprises the following contents:

research foundation

1. Refueling core fuel management strategy

The Haiyang nuclear power plant is an AP1000 type nuclear power plant, and currently adopts a reactor core fuel management strategy of a fixed period (18-month refueling period), low leakage and high fuel consumption. According to current fuel management strategies, there are 157 groups of fuel assemblies in the core, and the number of refueling assemblies per cycle is 64 groups. According to the currently adopted core fuel management strategy, the cycle lengths of the first cycle, the 2 nd cycle and the 3 rd cycle (balance cycle) of the AP1000 unit are respectively about 465EFPD (equivalent full power days), 510EFPD and 510EFPD, and the cycle lengths are relatively fixed. At the end of each cycle life, the unit will discharge all 157 sets of fuel, then replace the fuel assemblies with the deeper burn-up with 64 sets of fresh fuel assemblies with higher enrichment, and reload the 157 sets of fuel assemblies (including 64 sets of fresh fuel assemblies) after replacement into the core for the next cycle of operation.

2. Current core design for Haiyang nuclear power plant

The reactor core of the Haiyang nuclear power plant is loaded with 157 groups of AP1000 type fuel assemblies, the height of an active section is 4267.2mm, the thermal power of the reactor core is 3400MWt, and main design parameters related to the reactor core are given in table 1.

TABLE 1

Design parameter name	Design parameter size
		Electric Power, MW (e)	1250
RCS pressure, MPa (abs)	15.5
		Number of RCS loops	2
NSSS thermal power, MW (t)	3415
		Core thermal power, MW (t)	3400
Fuel assembly type	AP1000 type fuel assembly
		Number of fuel assemblies in reactor core	157
The heat release fraction in the fuel%	97.4
		Core active zone height (cold state), mm	4267.2
Core bypass flow%	5.9
		Inlet temperature (best estimated flow, 0 plugging pipe) of reactor core with rated power and DEG C	280.7
Rated power core outlet temperature (best estimated flow, 0 plugging tube) deg.C	323.3
		Equivalent diameter of reactor core, mm	3040.4
Height/diameter ratio of reactor core	1.40
		H2O/U molecular ratio, cell, cold state	2.40

The AP1000 fuel assembly consists of 264 fuel rods, 24 control rod guide tubes and 1 instrumentation tube, and is arranged in a 17X 17 square grid. Fuel rod made of UO₂The pellets are arranged in a ZIRLO alloy tube, and in order to reduce neutron leakage and improve the utilization rate of fuel, columnar or annular axial low-enrichment pellets are used at two ends of a fuel rod with higher enrichment. The material of the control rod guide tube and the instrument tube is ZIRLO alloy. Each fuel assembly has 8 middle location grids, 4 middle mixing grids, 1 top grid, 1 bottom grid and 1 protective grid, the middle location grids and the middle mixing gridsThe materials are ZIRLO alloy, and the materials of the top grillwork, the bottom grillwork and the protective grillwork are Inconel 718.

In order to control the core power distribution and to ensure a negative moderator temperature coefficient under power operating conditions, in the core design of a Haiyang nuclear power plant, an integral fuel burnable poison (IFBA) is used, which is a very thin layer of ZrB coated on the surface of the fuel pellets₂As a flammable absorber.

The total number of the reactor cores is 69 bundles of control rod assemblies, wherein 16 bundles of control rod assemblies are gray rod assemblies, and the absorber adopts tungsten (W); the 53 bundles are black rod assemblies, and the absorber adopts silver indium cadmium (Ag-In-Cd). The 69 bundles of control rods are divided into regulating rods and shutdown rods according to control functions. The adjusting rods are 37 bundles in total and comprise M rod groups and AO rod groups. The M rod components are MA, MB, MC, MD, M1 and M2 rod components, which are gray rod components with small reactivity value, and are inserted into or lifted out of the core according to a preset overlapping program to maintain the average temperature of the coolant to be changed according to a preset average temperature program. The AO rod set is mainly used for controlling the axial power distribution of the core in a proper range. The M rod set and the AO rod set complete the control of the reactor core reactivity and the power distribution in the MSHIM mode through two sets of independent controllers, so that the Haiyang nuclear power plant can realize the basic load operation of adjusting boron by stages and the load following operation without adjusting boron. The shutdown rods are 32 bundles, comprise SD1, SD2, SD3 and SD4 rod groups and are mainly used for shutdown. During shutdown, the regulating rod set and the shutdown rod set are all inserted into the core to provide enough shutdown allowance.

The second cycle of the Haiyang nuclear power plant employs a low leakage arrangement, employing only IFBA, and not WABA or other burnable poison. The number of the refueling assemblies was 64 groups and the two batches were divided by enrichment, with the average enrichment of the refueling assemblies (excluding the axial low enrichment zone) for the second cycle being 4.78% and 4.95%, respectively, and the number of assemblies being 36 and 28, respectively. The loading scheme for the second cycle core is shown in FIG. 1, wherein each square represents a group of fuel assemblies; the square box with the upper label "FFF" represents the new fuel assembly, and the corresponding lower label "XN _ YYY" indicates the original batch number of the new fuel assembly is X and the enrichment label is N, the number of integral burnable poison IFBA is YYY; the square with the upper reference numeral "XNN" represents the old fuel assembly and indicates the position of the old fuel assembly in the core of the previous cycle, X is a letter indicating the abscissa, NN is a number indicating the ordinate, and the corresponding lower reference numeral "XN" indicates the original batch number of the old fuel assembly as X and the enrichment index as N; the fuel assembly loading and unloading for the refueling cycle corresponding to the second cycle is shown in table 5. As can be seen from fig. 1, the second cycle of operation was loaded with fuel assemblies in a low leakage arrangement, with a portion of the old fuel assemblies being placed in the outermost turn, 28 new fuel assemblies of original lot number F and enrichment designation 2 (4.95%) being placed in the next outer turn, 36 new fuel assemblies of original lot number F and enrichment designation 1 (4.78%) and another portion of the old fuel assemblies being checkerboard-like in the interior region.

(II) technical problem

1. Poor flexibility in adjusting the shutdown time

At present, the reactor shutdown adjusting time of the Haiyang nuclear power plant can only be determined by extending operation and a reactor shutdown mode in advance, the adjustable window of the reactor shutdown time is small, the next circulation effect can be influenced by the extending operation, and the waste of fuel assemblies can be caused by the reactor shutdown in advance.

2. Poor economy

The longer the full power operation life in a single cycle of a nuclear power plant, the better the overall economy and the benefits in unit time. Currently, the haiyang nuclear power plant has realized 18 months refueling and still has the possibility of further prolonging the single cycle life.

(III) technical scheme

1. Summary of the technical solution

In order to increase the selectivity of the shutdown time of each cycle of the Haiyang nuclear power plant and improve the economy of the nuclear power plant, the flexible cycle refueling can be realized by increasing or reducing the number of refueling new fuel assemblies under the condition of design criteria and safety analysis permission. As the reactor cores of the pressurized water reactor nuclear power plant are basically arranged symmetrically by 1/4, the number of the fuel assemblies is adjusted to be multiples of 4, such as plus or minus 4 groups or plus or minus 8 groups.

By adopting a flexible circulation refueling scheme, the design and safety analysis influence of the refueling reactor core needs to be considered comprehensively. The specific steps for realizing the flexible circulating refueling scheme are as follows:

(1) determining an adjustable cycle length that may be needed

For plus or minus 4 groups (namely n is 1, 4n is 4), the adjustable cycle length comprises 60 groups, 64 groups and 68 groups on the basis of the original Haiyang nuclear power 64 groups new fuel assembly refueling scheme. For plus or minus 8 groups (namely n is 2, 4n is 8), the adjustable cycle length comprises 56 groups, 64 groups and 72 groups on the basis of the original Haiyang nuclear power 64 groups new fuel assembly refueling scheme.

(2) Determining flexible cycle refueling patterns

And adopting an analysis method under a limit working condition, such as maximum alternation, continuous maximum, continuous minimum, minimum after continuous maximum and maximum after continuous minimum. The sea Yang nuclear power plant adds or subtracts 4 groups of new fuel assembly refueling modes on the basis of the original 64 groups of new fuel assembly refueling and shows a table 2, the sea Yang nuclear power plant adds or subtracts 8 groups of new fuel assembly refueling modes on the basis of the original 64 groups of new fuel assembly refueling and shows a table 3, and the number of other adding or subtracting assemblies is analogized.

TABLE 2

Mode of reloading	The Nth cycle	Cycle N +1	Cycle N +2
				Maximum alternation	68	60	68
Continuous maximum	68	68	68
				Continuous minimum	60	60	60
Successive maximum and minimum	68	68	60
				Successive min to max	60	60	68

TABLE 3

Mode(s)	The Nth cycle	Cycle N +1	Cycle N +2
				Maximum alternation	72	56	72
Continuous maximum	72	72	72
				Continuous minimum	56	56	56
Successive maximum and minimum	72	72	56
				Successive min to max	56	56	72

(3) Performing tunable cycle design analysis

For all of the modes of tables 2 and 3, an adjustable cycle design analysis was performed to analyze the aspects that may be affected. Impact analysis includes, but is not limited to, the following:

a) the method comprises the steps of carrying out influence analysis on common key safety parameters of the reactor core of the nuclear power plant, wherein the parameters comprise reactor core operating power, flow, temperature, pressure, burnup, temperature coefficient, delayed neutron share, shutdown allowance, boron differential value, power factor and the like.

b) Analyzing the influence of accident analysis conclusion of the nuclear power plant. Full accident analysis or specific reactive accident analysis is adopted. The whole accident analysis is to re-analyze all design basis accidents in the final safety analysis report of the nuclear power plant; the specific reactivity accident analysis is to analyze the related accidents of reactivity, including boron dilution accident, control rod lifting accident, control rod step-out accident, rod ejection accident, steam pipeline fracture accident, loss of water accident, etc.

c) Functional and performance impact analysis is performed on nuclear power plant structures, systems and equipment.

d) And (4) analyzing the influence of the original radiation protection measures in normal operation and material change overhaul.

e) And (3) analyzing the influence of primary loop water chemistry of the nuclear power plant, and determining whether relevant chemical indexes can influence fuel performance or system structure.

f) And analyzing the influence of the decay heat of the spent fuel assembly.

g) And (4) analyzing a source item. And (3) analyzing the influence of the analysis results of source items such as the reactor core radioactivity accumulation amount, the primary circuit design reference source item, the shielding design equipment source item and the like of the original design scheme, and determining the influence on the capacity design and the shielding design of the spent fuel pool, the radioactive waste treatment system and other systems.

h) And analyzing the influence of boron crystallization after the loss of coolant accident. It is clear whether boron crystallization occurs after a loss of coolant accident.

i) And (4) carrying out influence analysis on the regular test execution condition and the core supervision condition in the technical specification. And (3) determining the influence on the periodic test execution of the unit.

(4) And providing reports such as final refueling design, thermal hydraulic design, reactor core operation limit value report, safety analysis and the like. The report comprises the contents of a refueling core design conclusion, unit starting and operating parameters, influence analysis conclusion, safety analysis conclusion and the like, and is used for nuclear power unit operation support and evidence obtaining.

(5) For the overall flexible cycle project, overall demonstration is required; and (4) adding or subtracting a certain number of fuel assemblies for a single cycle, and performing single cycle demonstration.

2. Integrated flexible cycle of Haiyang nuclear power plant

2.1 demonstration of the method

Aiming at the flexible circulation operation requirement, in order to demonstrate a mode of flexibly switching 64 +/-4 groups of refueling assemblies, namely the flexible switching among 60 groups of refueling new fuel assemblies, 64 groups of refueling new fuel assemblies and 68 groups of refueling new fuel assemblies, a disturbance mode of continuously and alternately adding or subtracting 4 groups of new fuel assemblies with the largest disturbance on the current 64 groups of refueling schemes is adopted, and the design and safety analysis of a reactor core fuel management scheme are carried out. FIG. 2 shows a flexible cycle fuel management strategy. The core fuel management mode of continuously and alternately adding and subtracting 4 groups of new assemblies can introduce disturbance to the current 64 groups of refueling modes to the maximum extent, so that the safety and feasibility of adding and subtracting 4 groups of refueling assemblies are evaluated conservatively.

The demonstration of the flexible cycle refueling is based on the existing safety evaluation standard (final safety analysis report FSAR), and complete demonstration analysis is developed from the aspects of the core cycle life, the fuel economy, the operation flexibility and the like, wherein the demonstration contents comprise the scheme arrangement of the flexible cycle refueling core, the analysis of the core general safety analysis parameters and the verification of relevant design criteria, the demonstration of power capacity, the verification of key accident envelope limit values and other aspects which are possibly influenced.

Based on reactor core fuel management reports of the Haiyang nuclear power 1 and 2 units, 4 groups of new fuel assemblies are continuously and alternately increased and decreased from the third fuel cycle to be stacked, namely, a reactor core scheme design that the number of the new fuel assemblies is respectively 68 and 60 groups, and the fuel replacement is alternately transited to a balance cycle is adopted. The enrichment of the refueling assembly, and the type of burnable poison, all follow the assembly type for the current 64-pack refueling management scheme balancing cycle, as shown in table 4 below.

TABLE 4

2.2 Flexible cycle refueling core design

2.2.1 design criteria and requirements

The main objectives of core nuclear design are: the rated thermal power output of 3400MWt is met under the condition of ensuring the safety of the reactor core; providing the physical characteristic parameters of the reactor core and the parameters and analysis results required by the safe startup, operation and shutdown of the corresponding reactor core.

The main design criteria and design requirements for core design are as follows:

(1) nuclear enthalpy rising thermal tube factor under thermal state full power condition

(2) Total power peak factor F under thermal state full power condition_QLess than or equal to 2.60 so as to meet the LOCA accident consequence limiting requirement;

(3) when the reactor core operates at any power level (including HZP), the temperature coefficient of the moderator cannot be a positive value, namely MTC is less than or equal to 0 pcm/DEG C;

(4) the reactor core is operated at any power level (including HZP), the shutdown margin (SDM) is more than or equal to 1600 pcm;

(5) the maximum average fuel consumption of the reactor core fuel rod does not exceed 62000MWd/tU so as to meet the requirement of fuel rod performance analysis.

2.2.2 design method and design program

While this report uses core computer program ANC for computational analysis, any other validated version of the available core design program can be used for flexible cycle analysis.

2.2.3 Flexible cycle core design

The flexible circulation (including the transition circulation and the balance circulation) adopts a low-leakage material distribution mode and only adopts IFBA. The refueling cycles corresponding to 68 groups of new fuel assemblies divide the refueling assemblies into two batches according to enrichment, the average enrichment (excluding the axial low-enrichment area) of the refueling assemblies is 4.45 percent and 4.95 percent respectively, and the number of the assemblies is 36 groups and 32 groups respectively. The refueling cycles corresponding to 60 groups of new fuel assemblies divide the refueling assemblies into two batches according to enrichment, the average enrichment (excluding the axial low-enrichment area) is 4.45 percent and 4.95 percent respectively, and the number of assemblies is 36 groups and 24 groups respectively.

And axial low-enrichment areas with 3.20% of enrichment degree are adopted at the top and the bottom ends of all the refueling assemblies. In order to reduce neutron leakage at two ends of the core active area, an axial low-enrichment area with the enrichment degree of 3.20% is adopted for all refueling assemblies. The fuel pellets with a height of 101.6mm at both ends of the non-low enrichment region of the active region are free of IFBA to flatten the core axial power distribution. And the IFBA is adopted to flatten the power distribution of the core and control the temperature coefficient of the moderator.

The calculation of the main physical parameters and the verification of the relevant design criteria were carried out on the core loading scheme.

Fig. 3 to 8 show core layouts of the third to eighth cycles, respectively, i.e., the transition to equilibrium cycles, wherein each square represents a group of fuel assemblies, the reference numerals of fig. 3 to 8 have the same meanings as those of fig. 2, and the square with the central center of the core, the upper reference numeral "XNN^*"indicates that the assembly is a first cycle discharge assembly and indicates the location of the old fuel assembly in the first cycle core, X is a letter indicating the abscissa and NN is a number indicating the ordinate; and the corresponding lower designation "XN" indicates that the original lot number of the old fuel assembly is X and the enrichment designation is N.

Table 5 shows the fuel assembly loading and unloading for the second through eighth cycles, and the subsequent alternating 68 sets of refueling cycles corresponding to 60 sets.

TABLE 5

Note: (1) means not including an axially low enrichment zone

Taking the third and fourth cycles and the subsequent alternating 68 and 60 refueling cycles, cycles N-1 and N, as an example, table 6 shows the average enrichment (excluding the axial low-enrichment zone) and burnable poison loading for each new batch of fuel assemblies.

TABLE 6

Note:

(1) y, Z denotes the batch number of the fuel assemblies of the subsequent batch after the third cycle, the same below;

(2) this means that the axial low enrichment zone is excluded, the same applies hereinafter.

Table 7 gives the main physical parameters of the third to eighth cycles.

TABLE 7

The fuel consumption statistics of the batches of fuel assemblies from the third cycle to the eighth cycle are summarized in tables 8 to 13, respectively. It should be noted that: in the main physical parameter table, the initial soluble boron concentration refers to the calculated values of BOL, HFP, NOXE, under critical conditions; maximum heat flux density heat pipe factorAnd maximum nuclear enthalpy rising thermal pipe factorRefers to the maximum calculated value under the whole fuel cycle, HFP, critical conditions; the cycle length was calculated under EOL, HFP, ARO, EQXE conditions with the critical soluble boron concentration controlled at 10 ppm.

TABLE 8

TABLE 9

Watch 10

TABLE 11

TABLE 12

Watch 13

2.2.4 Key safety parameter verification

On the basis of scheme design, the method carries out preliminary verification on key safety parameters such as reactor core power peak factors, fuel consumption limit values, moderator temperature coefficients, shutdown allowance and the like. From the results in Table 7 in section 2.2.3: transition cycle to equilibrium cycleThe maximum value is 1.503. after 8% uncertainty is considered,still less than 1.72;the maximum value is 1.797%, considering 8.15%After a total uncertainty of (including 5% of the calculated uncertainty, 3% of the engineering heat pipe factor, and 5.6% of the rod bend factor), the maximum heat flow density heat pipe factor for each cycle under HFP normal operating conditions does not exceed 2.60, meeting the relevant criteria requirements. The fuel consumption of the maximum fuel rod of each cycle is lower than 62000MWd/tU, and the fuel consumption design limit requirement is met. The service life of 68 groups of the refueling balance cycles is about 535.5EFPD, and the service life of 60 groups of the refueling balance cycles is about 476.2EFPD, so that the design requirement is met.

Table 14 shows the Moderator Temperature Coefficients (MTC) for the transition and equilibrium cycles versus operating conditions. Moderator temperature calculations were the first to analyze the most conservative HZP ARO NOXE regime that does not occur during plant operation. For the condition that the temperature coefficient of the moderator exceeds the limit value, the temperature coefficient of the moderator is verified by simulating the actual stack starting process of the nuclear power plant, and although the most positive MTC in the HZP ARO NOXE service life of 68 groups of refueling cycles is slightly larger than 0, the MTC in the stack starting process is smaller than 0, so that the MTC value can meet the related limit value requirement.

TABLE 14

Note: (1) means that the fuel cycle with the most positive MTC in the service life of HZP ARO NOXE is less than 0, and the verification of the most positive MTC in the service life in the starting process is not needed

Table 15 shows the hot trip margin calculations for the transition cycle and the equilibrium cycle, both meeting the limit requirement for trip margin not less than 1600 pcm.

Watch 15

Note: (1) it means that the design reference minimum trip margin is 1.60% Δ ρ.

By adopting an ANC program, 68 groups and 60 groups of alternating refueling core fuel management strategies are researched on the basis of the second fuel circulation of the core fuel management scheme of the unit of Haiyang nuclear power No. 1 and No. 2, and 68 groups and 60 groups of alternating refueling core fuel loading schemes are designed. On the basis, the calculation of main physical parameters and the verification of related design criteria are carried out on each cycle. The results and conclusions of the analysis were calculated as follows:

the subsequent cycles (including the transition cycle and the equilibrium cycle) also adopt a low-leakage distribution mode. 68 groups of the refueling circulation refueling assemblies are divided into two batches according to the enrichment degree, the average enrichment degree (excluding an axial low-enrichment area) of the refueling assemblies is 4.45 percent and 4.95 percent respectively, and the number of the assemblies is 36 groups and 32 groups respectively. The 60 groups of the refueling circulation refueling assemblies are divided into two batches according to the enrichment degree, the average enrichment degree (excluding an axial low-enrichment degree area) is 4.45 percent and 4.95 percent respectively, and the number of the assemblies is 36 groups and 24 groups respectively. And flattening the power distribution of the reactor core and controlling the temperature coefficient of the moderator by adopting IFBA burnable poison.

The service life of 68 groups of refueling balance cycles is about 535.5EFPD, the average unloading fuel consumption is about 50115MWd/tU, the average maximum rod fuel consumption is 59390MWd/tU, the service life of 60 groups of refueling balance cycles is about 476.2EFPD, the average unloading fuel consumption is about 49687MWd/tU, the average maximum rod fuel consumption is 57973MWd/tU, and the design requirements are met.

The design parameters from the transition cycle to the balance cycle all meet the requirements of the relevant design criteria, namely the nuclear enthalpy raising heat pipe factor, the heat flow density heat pipe factor, the moderator temperature coefficient and the maximum fuel rod burnup are all lower than the maximum limit value requirement, and the shutdown allowance is larger than the minimum limit value requirement.

In conclusion, the haiyang nuclear power plant has the capacity of alternately and flexibly reloading 68 groups and 60 groups, and the designed core scheme meets relevant design criteria and requirements.

3 Flexible cycle safety analysis

3.1 safety analysis parameters

(1) Computational analysis was performed on the general key safety parameters. Analysis shows that the new fuel assembly refueling scheme of the third cycle 68 of the unit No. 1 does not break through the general key safety parameters.

(2) Computational analysis was performed for reactive accidents. Analysis shows that except the boron dilution accident, other reactive accidents can not break through the safety analysis limit of the FSAR.

(3) The boron dilution accident calculation result shows that the initial boron concentration and the re-critical boron concentration in each mode of the transition circulation (68 groups of new materials) and the equilibrium circulation (68 groups of new materials) are higher than the current safety analysis limit value, so that the boron dilution accident reanalysis is carried out. The analysis results show that unexpected boron dilution during refueling is prevented by administrative intervention as specified by the technical specifications; unexpected boron dilutions during shutdown, startup, and power operations may be detected or trigger shutdown, and the operator has sufficient time to terminate boron dilution before re-criticizing.

3.2 Source term design and Radioactive Shielding calculation

The evaluation conclusion of the influence of the flexible cycle scheme design of the Haiyang nuclear power plant on the source item analysis and shielding design is as follows:

(1) the core radioactivity accumulation amount, the design reference reactor coolant source item, the emission source item, the shielding design equipment source item and other source item analyses of the original Haiyang nuclear power plant design scheme and the analysis result after the accident radioactivity have enough conservation, and the influence caused by the flexible refueling design scheme can be enveloped.

a) The result of the calculation of the reactor core radioactivity accumulation of the original Haiyang nuclear power plant design scheme takes the uncertainty caused by the uncertainty of power measurement and the change of the fuel management scheme into consideration, and is about 1.05. The original scheme result can envelop the influence brought by the design of the flexible loop scheme.

b) In the design benchmark reactor coolant fission product source item analysis of the original Haiyang nuclear power plant design scheme, in addition to considering the uncertainty, an uncertainty factor 1.15 generated by the normal purification flow change of the main loop water charge or the chemical and volume control system is also considered, and the influence caused by the flexible refueling scheme design can be enveloped; meanwhile, the calculation result of the neutron radiation field of the reactor core in the original design scheme is sufficiently conservative, and the calculation results of the N-16 source item and the activated corrosion product source item of the main loop are not influenced.

c) The total production of tritium in the primary loop is increased (less than 4.0%) due to the increase of the critical boron concentration of the core of 68 groups of new refueling cycles. From the perspective of emission source item analysis, the analysis result of the original design scheme is sufficiently conservative, and the increase of tritium generation amount caused by the increase of critical boron concentration can be enveloped.

d) The increase of the boron concentration of the main coolant can cause the waste liquid amount generated by boron regulation drainage to be slightly increased, but the waste liquid amount considered in the original emission source item analysis is sufficiently conservative, so the calculation result of the emission source item of the original design scheme can envelop the influence brought by the design of the flexible refueling scheme.

(2) The influence of the reactor core power distribution change caused by the flexible refueling scheme on the primary shielding calculation result is within 10 percent, and the primary shielding calculation analysis conclusion of the Haiyang nuclear power plant project cannot be broken through.

3.3 systematic evaluation and demonstration

Aiming at the design of a flexible circulation scheme of a Haiyang nuclear power plant, the influence of relevant systems or parameters of the power plant is evaluated.

(1) Chemistry of water

By adopting the flexible circulation scheme design, the upper limit of boron concentration of 68 groups of new fuel refueling circulation is slightly increased, but the primary circuit water chemistry still follows the control requirement of pHT6.9-7.4, and the Li concentration is not more than 3.5ppm under the condition of meeting the pHT during the power operation, thereby not influencing the fuel performance, not influencing the structural material of the system, and not changing the corresponding parameters of the water chemistry and the enveloping environment condition.

(2) Waste liquid discharge system

The increase of the boron concentration of the main coolant can lead to the small increase of the waste liquid amount generated by boron regulation drainage, but the design of the original system considers sufficient conservation and does not influence the capacity design conclusion of the three-waste treatment system.

(3) Decay heat

The design pair influence of the flexible circulation scheme is small, and the flexible circulation scheme can be enveloped by the calculation result of the decay heat of the design scheme of the original Haiyang nuclear power plant.

(4) System fuel assembly decay heat capability design and shielding design

The source item analysis and evaluation conclusion can show that the source item analysis results of the original Haiyang nuclear power plant design scheme, such as the reactor core radioactivity accumulation, the primary circuit design reference source item, the shielding design equipment source item, and the like, have sufficient conservation, and are not expected to influence the capability design and the shielding design conclusion of the spent fuel pool, the radioactive waste treatment system, and other systems.

(5) Boron crystallization after LOCA accident

For a LOCA accident, it is necessary to assess whether a change in boron concentration would result in boron crystallization after the LOCA accident. Because the water content and the boron concentration of the special system are obviously higher than the initial water content and the boron concentration in the reactor core, the special system has larger contribution to boron crystals, and the initial boron content of the reactor core has smaller contribution to the boron crystals. The optimization only improves the boron concentration in the main system, and the maximum increase amplitude of the core boron concentration can be enveloped by the original design analysis, so that the flexible refueling scheme can not cause the boron crystallization phenomenon after the LOCA accident.

(6) Evaluation of equipment overhaul period

The FSAR requires attention to the periodic tests specified in chapter 16 specifications. For example, if a periodic test requires a period of 18 months and the test is to be performed during a major repair, in which case the fuel cycle length is adjusted to 19 months, the performance of the test will be affected. According to the investigation, the technical specification does not have the regular test corresponding to the specific condition, so that the refueling period is adjusted from 18 months to 17-19 months, and the execution of each regular test in the technical specification is not influenced.

3.4 conclusion

The flexible cycle operation safety analysis of the Haiyang nuclear power plant shows that the design of each cycle reactor core meets the requirements of the final safety analysis limit value and the relevant accident acceptance criterion. The source item and the radioactive shielding calculation analysis can be enveloped by the original Haiyang nuclear power plant design scheme after evaluation, and design conclusion of relevant systems or parameters of the power plant of the original Haiyang nuclear power plant design scheme can not be influenced.

According to the technical scheme, on the basis of 64 groups of refueling design of an original Haiyang nuclear power plant, the number of 4 groups of refueling fuel assemblies is increased or reduced, and the design and safety analysis of the circulating refueling reactor core are carried out on the basis, so that the adjustability of the circulating length is finally realized, and the economy and the flexibility of the nuclear power plant are increased. Moreover, the technical scheme of the embodiment further improves the number adjustment of refueling fuel assemblies and the core loading arrangement, and performs refueling core design on the basis.

At present, the Haiyang nuclear power plant has completed the design of adding 4 groups of new fuel assemblies to the No. 1 unit and safety analysis and demonstration, and the third cycle of the No. 1 unit is about to implement the addition of 4 groups of new fuel assemblies for refueling. After 68 new fuel assemblies of the unit No. 1 are reloaded, powerful guarantee is provided for the nuclear energy heat supply of the Haiyang nuclear power in 2023 years. After the subsequent flexible circulation refueling project is integrally implemented, the flexible selection of the overhaul time can be realized, and support is provided for guaranteeing heating, power utilization low peak overhaul and the overhaul time adjustment of multiple units.

In conclusion, the novel method for managing the fuel in the reactor core of the pressurized water reactor can flexibly adjust the circulation length, solves the problems of poor flexibility of adjusting the shutdown time, poor economy and the like caused by relatively fixed circulation length, can adjust the original fixed refueling period of 18 months to 17-19 months, even 16-20 months, does not influence the execution of each periodic test in the technical specification, and can play a certain reference role for other nuclear power plants.

The applicant declares that the present invention illustrates the detailed structural features of the present invention through the above embodiments, but the present invention is not limited to the above detailed structural features, that is, it does not mean that the present invention must be implemented depending on the above detailed structural features. It should be understood by those skilled in the art that any modifications of the present invention, equivalent substitutions of selected components of the present invention, additions of auxiliary components, selection of specific modes, etc., are within the scope and disclosure of the present invention.

The preferred embodiments of the present invention have been described in detail, however, the present invention is not limited to the specific details of the above embodiments, and various simple modifications may be made to the technical solution of the present invention within the technical idea of the present invention, and these simple modifications are within the protective scope of the present invention.

It should be noted that the various technical features described in the above embodiments can be combined in any suitable manner without contradiction, and the invention is not described in any way for the possible combinations in order to avoid unnecessary repetition.

In addition, any combination of the various embodiments of the present invention is also possible, and the same should be considered as the disclosure of the present invention as long as it does not depart from the spirit of the present invention.