燃耗截面基本库对输运燃耗耦合计算的影响分析

doi:10.11889/j.0253-3219.2020.hjs.43.040004

Nuclear Techniques ›› 2020, Vol. 43 ›› Issue (4): 40004-040004.doi: 10.11889/j.0253-3219.2020.hjs.43.040004

• SPECIAL SECTION ON THE 11TH NATIONAL CONFERENCE ON NEW AND RESEARCH REACTORS (PART I) • Previous Articles Next Articles

Impact analysis of basic burnup cross section on coupled calculation of neutron transport burnup

Wankui YANG,Baoxin YUAN,Huan HUANG,Guanbo WANG,Songbao ZHANG

Institute of Nuclear Physics & Chemistry, China Academy of Engineering Physics, Mianyang 621900, China

Received:2020-01-16 Revised:2020-02-22 Online:2020-04-15 Published:2020-04-20
About author:YANG Wankui, male, born in 1988, graduated from China Academy of Engineering Physics with a master's degree in 2013, focusing on reactor physics and high performance computing

Abstract

Abstract: Background

Monte Carlo depletion method is based on Monte Carlo neutron transport code and ORIGEN2 point depletion code, it is widely used for neutron transport burnup. However, the continuous energy neutron cross sections of the nuclides in the current evaluation library are less than the cross section used in the depletion calculation. Therefore, the depletion cross sections generated by Monte Carlo transport calculation cannot replace all the cross sections in the basic depletion library.

Purpose

This study aims to analyse the influence of different basic depletion libraries on coupled calculation of neutron transport burnup.

Methods

The Monte Carlo transport depletion code MCBMPI (Monte Carlo Burnup code in MPI version) was employed to calculate the new VERA (Virtual Environment for Reactor Applications) depletion benchmark. The influence of different basic burnup cross section on the burnup calculation in transportation was compared and analyzed.

Results

Calculation results show that relative error of k_eff is less than 8‰ in all the depletion stages, and the mass variation of ²³⁵U and ¹³⁵Xe are less than 4‰ and 5‰ respectively, in the last depletion stage. The variation between thermal and fast neutron spectrum cross section library is small in this benchmark.

Conclusions

It is recommended to select a basic depletion cross section library containing similar neutron spectrum instead of a typical thermal neutron cross section library in real application.

Key words: Depletion cross section, Monte Carlo depletion, MCBMPI, VERA benchmark

CLC Number:

TL99

Wankui YANG, Baoxin YUAN, Huan HUANG, Guanbo WANG, Songbao ZHANG. Impact analysis of basic burnup cross section on coupled calculation of neutron transport burnup[J].Nuclear Techniques, 2020, 43(4): 40004-040004.

Figures/Tables 11

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Table 1

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Table 2

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Table 3

References 10

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项目Item	参数名称Parameter	参数值Value
燃料Fuel	材料Material	UO₂(ε=3.1%)
燃料Fuel	密度Density / g·cm^-3	10.257
	芯块半径Pellet radius / cm	0.409 6
	温度Temperature / K	900
包壳Cladding	材料Material	Zircaloy-4
包壳Cladding	密度Density / g·cm^-3	6.560
	内半径Inner radius / cm	0.418
	外半径Outer radius / cm	0.475
	温度Temperature / K	600
慢化剂Moderator	材料Material	硼酸Boric acid
慢化剂Moderator	密度Density / g·cm^-3	0.700
	温度Temperature / K	600

步长Step	燃耗Burnup / MW·d^-1·kg^-1	步长Step	燃耗Burnup / MW·d^-1·kg^-1	步长Step	燃耗Burnup / MW·d^-1·kg^-1
1	0.000 034 7	23	4.0	45	47.5
2	0.017 7	24	5.0	46	50.0
3	0.035 0	25	6.0	47	52.5
4	0.052 3	26	7.0	48	55.0
5	0.069 7	27	8.0	49	57.5
6	0.087 0	28	9.0	50	60.0
7	0.012 2	29	10.0	51	62.5
8	0.015 6	30	11.0	52	65.0
9	0.019 1	31	12.5	53	67.5
10	0.022 6	32	15.0	54	70.0
11	0.026 0	33	17.5	55	72.5
12	0.033 0	34	20.0	56	75.0
13	0.039 9	35	22.5	57	77.5
14	0.046 8	36	25.0	58	80.0
15	0.053 8	37	27.5	59	82.5
16	0.060 7	38	30.0	60	85.0
17	0.078 0	39	32.5	61	87.5
18	0.1	40	35.0	62	90.0
19	0.5	41	37.5	63	92.5
20	1.0	42	40.0	64	95.0
21	2.0	43	42.5	65	97.5
22	3.0	44	45.0	66	100.0

锕系元素 Actinides				裂变产物 Fission Products
1	²⁰⁶Pb	49	²⁴⁵Cm	1	³H	49	⁹⁷Mo	97	¹²⁴Sn	145	¹⁴³Ce
2	²⁰⁷Pb	50	²⁴⁶Cm	2	⁶Li	50	⁹⁸Mo	98	¹²⁴Sb	146	¹⁴³Pr
3	²⁰⁸Pb	51	²⁴⁷Cm	3	⁷Li	51	⁹⁹Mo	99	¹²⁴Te	147	¹⁴³Nd
4	²⁰⁹Bi	52	²⁴⁸Cm	4	⁹Be	52	⁹⁹Tc	100	¹²⁵Sn	148	¹⁴⁴Ce
5	²²³Ra	53	²⁴⁹Cm	5	⁶⁹Ga	53	⁹⁹Ru	101	¹²⁵Sb	149	¹⁴⁴Nd
6	²²⁴Ra	54	²⁵⁰Cm	6	⁷¹Ga	54	¹⁰⁰Mo	102	¹²⁵Te	150	¹⁴⁵Nd
7	²²⁶Ra	55	²⁴⁹Bk	7	⁷²Ge	55	¹⁰⁰Ru	103	¹²⁶Sn	151	¹⁴⁶Nd
8	²²⁷Ac	56	²⁵⁰Bk	8	⁷³Ge	56	¹⁰¹Ru	104	¹²⁶Sb	152	¹⁴⁷Nd
9	²²⁷Th	57	²⁴⁹Cf	9	⁷⁴Ge	57	¹⁰²Ru	105	¹²⁶Te	153	¹⁴⁷Pm
10	²²⁸Th	58	²⁵⁰Cf	10	⁷⁵As	58	¹⁰³Ru	106	¹²⁷Te	154	¹⁴⁷Sm
11	²²⁹Th	59	²⁵¹Cf	11	⁷⁶Ge	59	¹⁰³Rh	107	¹²⁷I	155	¹⁴⁸Nd
12	²³⁰Th	60	²⁵²Cf	12	⁷⁶Se	60	¹⁰⁴Ru	108	¹²⁸Te	156	¹⁴⁸Pm
13	²³²Th	61	²⁵³Cf	13	⁷⁷Se	61	¹⁰⁴Pd	109	¹²⁸Xe	157	¹⁴⁸Sm
14	²³³Th	62	²⁵⁴Cf	14	⁷⁸Se	62	¹⁰⁵Ru	110	¹²⁹Te	158	¹⁴⁹Pm
15	²³⁴Th			15	⁷⁹Se	63	¹⁰⁵Rh	111	¹²⁹I	159	¹⁴⁹Sm
16	²³¹Pa			16	⁷⁹Br	64	¹⁰⁵Pd	112	¹²⁹Xe	160	¹⁵⁰Nd
17	²³²Pa			17	⁸⁰Se	65	¹⁰⁶Ru	113	¹³⁰Te	161	¹⁵⁰Sm
18	²³³Pa			18	⁸⁰Kr	66	¹⁰⁶Pd	114	¹³⁰I	162	¹⁵¹Pm
19	²³²U			19	⁸¹Br	67	¹⁰⁷Pd	115	¹³⁰Xe	163	¹⁵¹Sm
20	²³³U			20	⁸²Se	68	¹⁰⁷Ag	116	¹³¹I	164	¹⁵¹Eu
21	²³⁴U			21	⁸²Kr	69	¹⁰⁸Pd	117	¹³¹Xe	165	¹⁵²Sm
22	²³⁵U			22	⁸³Kr	70	¹⁰⁸Cd	118	¹³²Te	166	¹⁵²Eu
23	²³⁶U			23	⁸⁴Kr	71	¹⁰⁹Ag	119	¹³²Xe	167	¹⁵²Gd
24	²³⁷U			24	⁸⁵Kr	72	¹¹⁰Pd	120	¹³³Xe	168	¹⁵³Sm
25	²³⁸U			25	⁸⁵Rb	73	¹¹⁰Ag	121	¹³³Cs	169	¹⁵³Eu
26	²³⁹U			26	⁸⁶Kr	74	¹¹⁰Cd	122	¹³⁴Xe	170	¹⁵⁴Sm
27	²⁴⁰U			27	⁸⁶Rb	75	¹¹¹Ag	123	¹³⁴Cs	171	¹⁵⁴Eu
28	²³⁵Np			28	⁸⁶Sr	76	¹¹¹Cd	124	¹³⁴Ba	172	¹⁵⁴Gd
29	²³⁶Np			29	⁸⁷Rb	77	¹¹²Cd	125	¹³⁵I	173	¹⁵⁵Eu
30	²³⁷Np			30	⁸⁷Sr	78	¹¹³Cd	126	¹³⁵Xe	174	¹⁵⁵Gd
31	²³⁸Np			31	⁸⁸Sr	79	¹¹³In	127	¹³⁵Cs	175	¹⁵⁶Eu
32	²³⁹Np			32	⁸⁹Sr	80	¹¹⁴Cd	128	¹³⁵Ba	176	¹⁵⁶Gd
33	²³⁶Pu			33	⁸⁹Y	81	¹¹⁴Sn	129	¹³⁶Xe	177	¹⁵⁷Eu
34	²³⁷Pu			34	⁹⁰Sr	82	¹¹⁵Cd	130	¹³⁶Cs	178	¹⁵⁷Gd
35	²³⁸Pu			35	⁹⁰Y	83	¹¹⁵In	131	¹³⁶Ba	179	¹⁵⁸Gd
36	²³⁹Pu			36	⁹⁰Zr	84	¹¹⁵Sn	132	¹³⁷Cs	180	¹⁵⁹Tb
37	²⁴⁰Pu			37	⁹¹Y	85	¹¹⁶Cd	133	¹³⁷Ba	181	¹⁶⁰Gd
38	²⁴¹Pu			38	⁹¹Zr	86	¹¹⁶Sn	134	¹³⁸Ba	182	¹⁶⁰Tb
39	²⁴²Pu			39	⁹²Zr	87	¹¹⁷Sn	135	¹³⁸La	183	¹⁶⁰Dy
40	²⁴³Pu			40	⁹³Zr	88	¹¹⁸Sn	136	¹³⁹La	184	¹⁶¹Dy
41	²⁴⁴Pu			41	⁹³Nb	89	¹¹⁹Sn	137	¹⁴⁰Ba	185	¹⁶²Dy
42	²⁴¹Am			42	⁹⁴Zr	90	¹²⁰Sn	138	¹⁴⁰La	186	¹⁶³Dy
43	²⁴²Am			43	⁹⁴Nb	91	¹²¹Sb	139	¹⁴⁰Ce	187	¹⁶⁴Dy
44	²⁴³Am			44	⁹⁵Zr	92	¹²²Sn	140	¹⁴¹Ce	188	¹⁶⁵Ho
45	²⁴⁴Am			45	⁹⁵Nb	93	¹²²Te	141	¹⁴¹Pr	189	¹⁶⁶Er
46	²⁴²Cm			46	⁹⁵Mo	94	¹²³Sn	142	¹⁴²Ce	190	¹⁶⁷Er
47	²⁴³Cm			47	⁹⁶Zr	95	¹²³Sb	143	¹⁴²Pr	191	¹⁶⁸Er
48	²⁴⁴Cm			48	⁹⁶Mo	96	¹²³Te	144	¹⁴²Nd	192	¹⁶⁹Tm

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Impact analysis of basic burnup cross section on coupled calculation of neutron transport burnup

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