ajith v d, david nesaraj, y.k.pandey, g. biswas nuclear ... on spent fuel... · ajith v d, david...

Determination of Fission Product Activity for Spent Fuel Characterization.

Ajith V D, David Nesaraj,Y.K.Pandey, G. Biswas

Nuclear Power Corporation of India

Ltd, India.

INDIAN NUCLEAR POWER CAPACITY

Reactors in Operation

Plant Name Reactor

Type

Start of

Commercial

Operation

Rated

Capacity(Mwe)

TAPS 1&2 BWR 1969,1969 2x160

TAPS 3&4 PHWR 2006,2005 2x540

RAPS 1&2 PHWR 1973,1981 100,200

RAPS 3&4 PHWR 2000,2000 2x220

RAPS 5&6 PHWR 2010,2010 2x220

MAPS 1&2 PHWR 1984,1986 2x220

NAPS 1&2 PHWR 1991,1992 2x220

KAPS 1&2 PHWR 1993,1995 2x220

KGS 1&2 PHWR 2000,2000 2x220

KGS 3&4 PHWR 2007,2011 2x220

KKNPP 1&2 VVER 2014,2017 2x1000

Total Nuclear Power Plant Capacity 6780

12–14 November, 2019

2Directorate of Engineering-LWR, NPCIL

Projects Under Construction

Plant Name Reactor

Type

Rated

Capacity(Mwe)

Expected Year

of Operation

KAPS 3&4 PHWR 2x700 2019,2020

RAPS 7&8 PHWR 2x700 2020,2021

KKNPP 3&4 VVER 2x1000 2023,2023

12–14 November, 2019 Directorate of Engineering-LWR, NPCIL3

PWR Operating Units in India

Two units of PWR( VVER-1000 types) at Kudankulam

Site

◦ Unit #1 Operating in its 4th Cycle

◦ Unit#2 Operating in its 2nd Cycle

One Year cycle lengths for the units


Fuel Types used

UTVS Design

Total 11 types of FAs viz.

16

24, 24B20, 24B36

36, 36B20, 36B36

40, 40B20, 40B36, 40B50

[ FA TYPES 36 USES 3.3% & 3.7% U-235]

[ FA TYPES 40 USES 3.7% & 4.1% U-235


Core Cross section view


0

5

10

15

20

25

30

35

40

0 10 20 30 40 50 60

Weig

thF

racti

on

(g/k

g-U

)

Burnup(MWD/KgU)

Weight Fractions with Burnup (36 Type)

U235

Total Pu

Fissile Pu


0

5

10

15

20

25

30

35

40

45

0 10 20 30 40 50 60

Weig

th F

racti

on

(g/k

g-U

)

Burnup(MWD/KgU)

Weight Fractions with Burnup (40 Type)

U235

Total Pu

Fissile Pu


Initial Core loading pattern

36 1

36 2

36 3

36 4

36 5

36 6

36 7

36B36 8

24 9

24B36 10

24 11

24B36 12

24 13

36B36 14

36 15

36 16

24 17

16 18

16 19

16 20

16 21

16 22

16 23

24 24

36 25

36 26

24B36 27

16 28

24B36 29

24B36 30

24 31

24B36 32

24B36 33

16 34

24B36 35

36 36

36 37

24 38

16 39

24B36 40

16 41

16 42

16 43

16 44

24B36 45

16 46

24 47

36 48

36 49

24B36 50

16 51

24 52

16 53

24 54

24B20 55

24 56

16 57

24 58

16 59

24B36 60

36 61

36 62

24 63

16 64

24B36 65

16 66

24B20 67

16 68

16 69

24B20 70

16 71

24B36 72

16 73

24 74

36 75

36B36 76

16 77

24B36 78

16 79

24 80

16 81

24 82

16 83

24 84

16 85

24B36 86

16 87

36B36 88

36 89

24 90

16 91

24B36 92

16 93

24B20 94

16 95

16 96

24B20 97

16 98

24B36 99

16100

24101

36102

36103

24B36104

16105

24106

16107

24108

24B20109

24110

16111

24112

16113

24B36114

36115

36116

24117

16118

24B36119

16120

16121

16122

16123

24B36124

16125

24126

36127

36128

24B36129

16130

24B36131

24B36132

24133

24B36134

24B36135

16136

24B36137

36138

36139

24140

16141

16142

16143

16144

16145

16146

24147

36148

36149

36B36150

24151

24B36152

24153

24B36154

24155

36B36156

36157

36158

36159

36160

36161

36162

36163


4th Cycle loading pattern(Unit-1)


Spent Fuel Storage pool(At Reactor)

Designed capacity for storage for a period of 8 years(+full core unloading at any moment of NPP operation)◦ 582 cells for FAs

◦ 64 cells for sealed cans

Database of the FA details in the pool is maintained◦ IDs can be read by Fueling Machine camera

Failed FAs determined by sipping of FAs in the can◦ Iodine content

◦ Cs content

Minimum storage of 5 years in the pool

Construction of Away-From-Reactor(AFR) facility is planned for interim storage


Spent Fuel in KKNPP

Spent FAs presently in Spent Fuel Pool( At Reactor)

Unit Cycle

Number of

FAs

discharged

Type

Average

Burnup

(MWD/KgU)

16 24 36

Unit#1

Cycle1 END 56 54 2 0 12.4

Cycle2 END 48 2 46 0 26.0

Cycle3 END 49 0 19 30 34.7

Unit#2 Cycle1 END 54 54 0 0 11.5

Total No of FAs in the

pool 207


Spent Fuel Characteristics Studied

Radiological characterization

Thermal characterization

Criticality


Radiological characteristics

Spent Fuel activity determination

◦ Fission products data

◦ Activation products data

◦ Actinides data

◦ The specific reactor conditions used for generating these data


Determination of Spent fuel Activity

Assessment of Radio isotope inventory of Core of KKNPP Units with UTVS type of fuel assemblies were performed

Isotopic inventory calculations were performed by MCU-REA Computer Code MCU-BURNUP

MCU/ORI module

Results of Radioisotope inventory for KKNPP Core(At the time of discharge) for enrichment 4.0% and for a burnup of 40 MWD/kgU are shown in the table below.


Assessment of core Radioisotopes inventory in VVER-1000

Isotope Activity(Bq)

Radio Activity of Inert gases

Kr-83M 3.18E+17

Kr-85 2.97E+16

Kr-85M 6.90E+17

Kr-87 1.31E+18

Kr-88 1.73E+18

Kr-89 2.14E+18

Xe-133M 1.82E+17

Xe-133 5.68E+18

Xe-135 1.55E+18

Xe-137 4.87E+18

Xe-138 4.47E+18


Radio Activity of volatile elements

H-3 1.60E+15

Rb-86 6.58E+15

Rb-88 1.76E+18

Rb-89 2.32E+18

Sb-127 2.84E+17

Sb-129 9.54E+17

Te-127 2.69E+17

Te-127M 4.63E+16

Te-129 8.92E+17

Te-129M 1.47E+17

Te-132 3.95E+18

I-131 2.79E+18

I-133 5.60E+18

I-134 6.14E+18

I-135 5.32E+18

Cs-134 5.62E+17

Cs-137 3.24E+17




Radio Activity of Intermediate

volatile elements

Sr-89 2.40E+18

Sr-90 2.27E+17

Sr-91 3.09E+18

Sr-92 3.36E+18

Ru-103 4.52E+18

Ru-105 3.23E+18

Ru-106 1.62E+18

Ba-139 5.05E+18

Ba-140 4.74E+18


Radio Activity of Refractory Elements

and Fuel Particles

Y-90 2.43E+17

Y-91 3.16E+18

Y-92 3.38E+18

Y-93 3.88E+18

Zr-95 4.47E+18

Zr-97 4.53E+18

Nb-95 4.47E+18

Rh-105 3.04E+18

Pd-107 3.19E+11

Pd-109 1.07E+18

La-141 4.45E+18

La-142 4.14E+18




Radio Activity of Refractory Elements

and Fuel Particles

Pr-143 4.12E+18

Ce-141 4.41E+18

Ce-143 4.12E+18

Ce-144 3.58E+18

Nd-147 1.77E+18

Am-241 3.93E+14

Am-243 6.27E+13

Cm-242 1.33E+17

Cm-244 7.19E+15

Np-239 6.05E+19

Pu-238 8.85E+15

Pu-241 4.21E+17


Decay Heat Calculations

Calculations performed for type of FAs at KKNPP

Methods Available

◦ ANSI/ANS-5.1-2005

◦ US-NRC Regulatory Guide 3.54(“SPENT FUEL HEAT GENERATION IN AN

INDEPENDENT SPENT FUEL STORAGE INSTALLATION”)

◦ MCU-REA1/ORI

12–14 November, 2019 Directorate of Engineering-LWR, NPCIL 19


ANSI/ANS-5.1-2005

1.0E-08

1.0E-07

1.0E-06

1.0E-05

1.0E-04

1.0E-03

1.0E-02

1.0E-01

1.0E+00

1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04 1.0E+05 1.0E+06 1.0E+07 1.0E+08 1.0E+09 1.0E+10

P/P

0(l

og

sc

ale

)

Time(sec-log scale)

Variation of P/Po with time after 4 cycles of operation of VVER-1000 Fuel

Code DECAYHEAT(ANS-5.1 ,U-235 only; R=1.0,w/o error)

Code DECAYHEAT(ANS-5.1,All nuclides, R=0.6,w/o error)



0

1

2

3

4

5

6

7

0 5 10 15 20 25 30 35 40 45 50

Decay H

eat(

KW

)

Time(Year)

Variation of Decay Heat(in KW) with time after 4 cycles of operation of VVER-1000 Fuel

Code DECAYHEAT(ANS-5.1 ,U-235 only; R=1.0,+3sigma err)

Code DECAYHEAT (RG-3.54), with error

Decay Power for VVER FA(US-NRC, RG 3.54 Methods)


Criticality

Criticality calculations were performed for

◦ Spent Storage fuel pool

◦ Proposed AFR Facility

Performed using MCU-REA code

No Burnup credit considered presently


Conclusions

Evaluation of Characteristics data of Spent VVER fuel

(Radiological, thermal & Criticality) were performed in

NPCIL and are available for further use in back-end

processes.


Thank You for your Kind attention.

24Directorate of Engineering-LWR, NPCIL12–14 November, 2019

ajith v d, david nesaraj, y.k.pandey, g. biswas nuclear ... on spent fuel... · ajith v d, david...

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