neutron activation of reactor components during …

1Second International Symposium on Nuclear Power Plant Life Management15-18 October 2007, Shanghai, China

NEUTRON ACTIVATION OF REACTOR COMPONENTS DURING OPERATION LIFETIME OF A NPP

Gunter Pretzsch, Bernhard Gmal, Ulrich Hesse and Klemens Hummelsheim

Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbHKurfürstendamm 200, 10719 Berlin, Germany

[email protected]


Content

� Introduction

� GRS code systems for activation analysis

� Validation against experimental data

� Comparison of a standard and an extended calculation method

� Example for application: PWR 40 years operation time


Introduction� Knowledge of the activation level of neutron irradiated materials during lifetime

of a NPP important for decommissioning� Evaluation of new decommission technologies with respect to radiation

protection and waste management� Besides measurements theoretical activation calculations provide important information, e.g. for work planning� Determination of inventories for management and disposal of the activated materials


GRS code systems for activation analysis

� Own-developed systems with the updated ORNL code ORIGEN as main tool� Standard system GRSAKTIV is based on ENDF/B-V data for three neutron

energy groups up to 10 MeV for six neutron reaction channels � Extended activation code GRS-ORIGENX is based on point data files ENDF/B-VI / JEF2.2 / JENDL3.2 / European Activation File 97, here

condensed to six groups up to 20 MeV for 15 channels� Application will be shown with respect to decommissioning of NPP� Extended neutron reaction channels and program flow chart in the next files


Development of advanced methods and libraries for activation calculations

Standard

Extended


1d ANITABLE

2d R-Z DORT

ABBRAND1d Burn-up calculation

ABSCHIRMUNGIsotopic XS, n-γ source terms,

2d radiation transport calculation, 2d dose

calculation, 2d neutron spectra calculation

AKTIVIERUNGActivation

calculations for multiple material

regions

1d OREST

DORTABLE GRSAKTIV

GRS-ORIGENX

Program Flow Chart of the burn-up, shielding and activation AAA-Sequence


Validation of the AAA-Sequence Activation Method� Calculation of 3 cycle irradiation of assembly structure materials in a full core

RZ DORT simulation and C/E comparison with experimental data (upper part, singular head screw) of a German NPP

� Calculation of 1 hour irradiation of a concrete sample of biological shield in a highly thermalized neutron flux and C/E comparison with experimental data of a Japanese research reactor

� C/E = Calculated / Experimental� Both methods GRSAKTIV and ORIGENX show satisfying results. Most

deviations are similar to other publications. But �� New ORIGENX inelastic scatter reaction (n,n1) for 93m Niobium can now be

compared to experiment � new:


Comparison of activation results GRSAKTIV and GRS-ORIGENX at a single head screw of assembly BE318 of NPP BIBLIS at discharge after 3 fuel cycles irradiation

00,10,20,30,40,50,60,70,80,91

1,11,21,31,41,51,61,71,8

71 d 5,3 a 312 d 84 d 2,7 a 45 d 28 d 8E4 a 100 a 2E4 a 16 a58Co 60Co 54Mn 46Sc 55Fe 59Fe 51Cr 59Ni 63Ni 94Nb 93mNb

Analyzed radioactive Isotopes in Steel <1.4541> and their Halflifes

Calcu

lated

/Expe

rimen

t

C/E GRSAKTIV

C/E ORIGENXnew:


Comparison of activation results GRSAKTIV and GRS-ORIGENX in concrete after 1 hour irradiation of JAERI JRR-4 research reactor

0,000,100,200,300,400,500,600,700,800,901,001,101,201,301,401,501,601,701,80

46Sc 51Cr 54Mn 59Fe 60Co 65Zn 124Sb 134Cs 141Ce 152Eu 177Lu 181Hf 182Ta 233PaAnalyzed radioactive Isotopes in Concrete

Calcu

lated

/Expe

rimen

t

C/E GRSAKTIV

C/E ORIGENX


Applications with regard to the operation lifetime of a real NPP� A: Core inventory and flux strength in the active region calculated by OREST� AA: Reactor simulation with assembly structures, core basket, reactor vessel

and biological shield and coolant with axially different density and temperature with DORT-DORTABLE in 83 groups, anisotropic order PL=3, 32 neutron energy groups for thermal up-scatter

� Automatically neutron spectra and attenuation factors are calculated in 187 material regions and more than 10,000 mesh points for activation calculations

� AAA: GRSAKTIV and ORIGENX activation calculations based on the operation lifetime 40 years of a real PWR with 3000 MW thermal energy output

� Next figure shows normalized flux distributions of DORT at core, vessel and biological shield (other regions omitted)


1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19

corecore basket

RV2RV5

Bio1Bio4

1,00E-08

1,00E-07

1,00E-06

1,00E-05

1,00E-04

1,00E-03

1,00E-02

1,00E-01

1,00E+00

1,00E+01

Norm

alize

d Neu

tron F

lux St

reng

th

Axial Layers

Radial Layers

coresurroundwater1core basketwater2RV1RV2RV3RV4RV5airisolationBio1Bio2Bio3Bio4Bio5

Lower and upper part of Core

Vessel

Concrete


Activation results of reactor vessel in the short time range one month after shut down and 40 years operation lifetime� Calculations of flux distributions of 83 groups DORTABLE repeated and proved

with three other deep penetration cross section libraries up to 175 groups (EURLIB-99)

� Neutron fluxes vary strongly inside of vessel material � five material layers RV1-RV5 of steel <22NiMoCr37> had to be activated separately and averaged

� Next table for reactor vessel steel shows excerpt of assumed trace elements <22NiMoCr37> and most important radioactive isotopes calculated with STAndard and EXTended Method GRSAKTIV (=ENDF/B-V) and ORIGENX (=ENDF/B-VI + others) in 83 energy groups

� Differences between STA and EXT not so great. Problems Co-58 and Mo-99!� New Isotope Fe-60 detected in calculation, but very low activities!


+41.16E+061.12E+06Total Activity�-64.49E+024.24E+02152Eu0.1 63-Eu-32.75E+012.66E+01134Cs0.1 55-Cs

+ 44 �7.49E+001.08E+0199Mo+ 201.40E+011.68E+0193Mo7,00042-Mo+156.88E+037.88E+0363Ni+86.42E+016.91E+0159Ni

+ 86 �4.31E+038.05E+0358Co9,00028-Ni+0.28.25E+048.27E+0460Co30027-Co

� New reaction5.52E-080.060Fe+ 271.90E+042.43E+0459Fe-59.67E+059.21E+0555Feremainder26-Fe-156.11E+045.22E+0454Mn10,00025-Mn / 26-Fe+161.41E+041.64E+0451Cr5,00024-Cr-81.99E+021.84E+0232P20015-P+41.37E+021.42E+0214C2,100 / 590 06-C / 07-N -13.25E+033.22E+033H1.0 03-Li

Difference %(STA-EXT)/EXT

GRS-ORIGENX EXTended (Bq/g)

GRSAKTIV STAndard (Bq/g)

Nuclideactivated

Content (ppm)TraceElements


Activation of the NPP biological shield in the short time range one month after shut down and 40 years operation lifetime

� Neutron fluxes vary very strongly inside of biological shield � five material layers BIO1-BIO5 of 92 % concrete with 8 % construction steel had to be separately activated and averaged

� Neutron spectra are very thermalized in contrast to the reactor vessel� Next table for the biological shield shows excerpt of trace elements and most

important radioactive isotopes calculated with STAndard and EXTendedMethod GRSAKTIV (=ENDF/B-V) and ORIGENX (=ENDF/B-VI + others)

� Differences between STA and EXT not so great. Problems Ar-37 and Co-58 !


-151.84E+41.56E+04Total Activity�-231.10E+018.53E+00154Eu07.78E+017.77E+01152Eu0.1363-Eu-282.54E+011.82E+01134Cs0.6155-Cs-93.62E+003.30E+0065Zn10.6930-Zn+ 27.15E+017.26E+0163Ni

+ 50 �5.05E+007.55E+0058Co9,00028-Ni-101.04E+039.43E+0260Co3002.3727-Co02.27E+022.27E+0259Fe-161.13E+049.47E+0355Feremainder10,50026-Fe-328.07E+015.52E+0154Mn10,00025-Mn-21.48E+021.45E+0251Cr50009.0724-Cr-36.95E+016.76E+0146Sc0.1 2.8121-Sc-14.05E+033.99E+0345Ca-32.24E+012.17E+0141Ca

-62 �1.29E+034.86E+0237Ar195,00020-Ca-33.39E+013.29E+013H1.0 03-Li

Difference %(STA-EXT)/EXT

GRS-ORIGENXEXTended (Bq/g)

GRSAKTIVSTAndard (Bq/g)

Nuclideactivated

8 %Steel (ppm)

92 %Concrete (ppm)

TraceElements


Analyses of important radioactive isotopes in the intermediate time range up to 100 years after shutdown of NPP

� Next figures show activities of most important isotopes in the reactor vessel and the biological shield in units (Bq/g)

� These calculations done by using the GRSAKTIV system� In the vessel the activities after one year cooling time are dominated by Fe-55,

Co-60, Ni-59 and Ni-63� In the biological shield plus construction steel the activities after one year

cooling time are dominated by C-14, Fe-55, Co-60, Ni-63 and Eu-152


Activities reactor vessel after 40 years operation time in units Bq/g

1,00E-051,00E-041,00E-031,00E-021,00E-011,00E+001,00E+011,00E+021,00E+031,00E+041,00E+051,00E+061,00E+07

0 10 20 30 40 50 60 70 80 90 100decay time

spec

ific ac

tivity

Bq/g

H 3C 14CR 51MN 54FE 55FE 59CO 58CO 60NI 59NI 63NB 93MNB 94MO 93CS134EU152totals

totals


Activities reactor biological shield after 40 years operation time in units Bq/g

1,00E-05

1,00E-04

1,00E-03

1,00E-02

1,00E-01

1,00E+00

1,00E+01

1,00E+02

1,00E+03

1,00E+04

1,00E+05

0 10 20 30 40 50 60 70 80 90 100decay time

spec

ific ac

tivity

Bq/g

H 3C 14CR 51MN 54FE 55FE 59CO 58CO 60NI 59NI 63NB 93MNB 94MO 99CS134EU152totals


Analyses of important Gamma emitting isotopes in the intermediate time range up to 100 years after shutdown of NPP� Many radioactive isotopes are weak gamma emitters (H-3, Fe-55), but gamma

emission is important for dose rates and radiation protection� Next figures shows Gamma power (W/g) of important isotopes in reactor vessel

and biological shield � The calculations are done by using GRS-ORIGENX� In the vessel the activities after one year cooling time are mostly dominated by

Co-60 and at 100 years cooling time additionally Eu152 and Ni-59� In the biological shield plus construction steel the activities after one year are

dominated by Co-60 and Eu-152� Cobalt contents in structure steel materials vary from ~100 up to ~ 3000 ppm!� Europium trace concentrations very low ~ 0.1 ppm and not always well known!


ORIGENX: Reactor Vessel Gamma Power (W/g) after 40 years operation time

1,00E-15

1,00E-14

1,00E-13

1,00E-12

1,00E-11

1,00E-10

1,00E-09

1,00E-08

1,00E-07

1,00E-06

0 10 20 30 40 50 60 70 80 90 100decay time

spec

ific G

amma

Powe

r W/g

MN 54

FE 59

CO 58

CO 60

NI 59

NB 93M

MO 93

EU152

EU154

EU155

totals


ORIGENX: Reactor Biological Shield Gamma Power (W/g) after 40 years operation time

1,00E-15

1,00E-14

1,00E-13

1,00E-12

1,00E-11

1,00E-10

1,00E-09

1,00E-08

0 10 20 30 40 50 60 70 80 90 100decay time

spec

ific G

amma

Powe

r W/g

SC 46

MN 54

FE 59

CO 58

CO 60

EU152

totals

Eu-152


Summary and conclusions� Satisfying validation of the GRS AAA Burn-up -, Shielding- and Activation-

Sequence with the main codes OREST, 2d DORT, GRSAKTIV and GRS-ORIGENX

� Tested against experiments of activated structure material steel and concrete� Application to operation life time up to 40 years PWR of 3000 MW thermal

power� Most important radioactive isotopes and gamma emitters analyzed up to

100 years cooling time after shutdown� Problematic isotopes (Co-58, Ar-37) and trace elements (Co, Eu) identified� Further full documented activation experiments and applications needed

neutron activation of reactor components during …

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