Nuclear Presentation

1. Development of computer program for
calculation of projected dose in the off-site
Area for Design Based and Beyond Design
Based Accidents at NPPs.
PRESENTED BY: TALHA JAVED
SUPERVISOR: Mr. QAISAR NADEEM
CO-SUPERVISOR: Dr. HASEEB-ur-REHMAN

2. MOTIVATION
 NPPs release radioactive effluents in environment under normal and emergency
situation.
 While atmospheric Releases In NPP’s emergency conditions:
 Projection of dose to offsite locations is quite important.
 Make decision based on the exposure of general public like:
 Evacuation.
 Shelter.
 Precautions etc.
 We need tools to project offsite doses and suggest required actions to Decision
Making Authorities.

3. SCOPE & OBJECTIVES
 EDSS is designed to:
 Predict offsite dose dispersion upto 25km radius due to radioactive effluent releases
specific to CHASNUPP site.
 Tabulated data of dose at required locations.
 Visual representation of dose dispersion and identifying the areas of high dose.
 Assist Decision Makers in making decision regarding general public.
 EDSS is designed to be used at CHASNUPP site and needs to be independent of
any internet facility usage.

4. WHY WE NEEDED TO DEVELOP EDSS
EDSS is a site specific tool(CHASNUPP) and we were called for help in developing a
software utility for predicting off-site doses due to atmospheric releases of
radioactive effluents.
Currently client is using INTERASS for this purpose which is based on Gaussian Plume
Model which has its inherent limitations like:
 Recommended for 10 km radius.
 Wind direction and speed is assumed to be constant along the track of plume.
 Recommended for wind velocities greater than 3 m/sec which is not the case for
CHASNUPP.

5. INTRODUCTION TO EDSS
 EDSS is a software utility for prediction offsite doses and is basically an integration of
different utilities performing various steps for predicting dose dispersion. The utilities
include:
 CALMET for generating 3D wind field based on Weather profiles and Geodetic data.
 LAPMOD (based on Langrangian Particulate model) and dose factors for calculation of dose
dispersion based on CALMET and Source data input.

6. INTRODUCTION TO EDSS
 Main Window
 EDSS requires following inputs for
dose calculations.
 Run period
 Meteorological data
 Source data

7. INTRODUCTION TO EDSS
 EDSS Calculates and stores following Doses
is separate files:
 Submersion Dose
 Deposition Dose
 Total Dose
 EDSS also provides visual representation of
dose dispersion in form of contours of
different doses.

8. My Tasks
Main Tasks:
 Theoretical Source Term estimation incorporation.
 Introduce Inverse Modeling in EDSS.
Additional Tasks:
 Making the EDSS’s visualization more better and user friendly.
 Develop argument based input mechanism.
 Making EDSS compatible for CHASNUPP.
 Develop users manual for EDSS.

9. SEVERE ACCIDENT SOURCE TERM
What is a severe accident?
 An accident In NPP for which it wasn’t designed i.e. beyond design based accidents.
What is source term?
 Source term consists of the information of:
 Type of source i.e. point source, line source etc.
 Emission parameters like:
 Type of radionuclides emitted.
 Activity of each radionuclide.
 Temperature and velocity of each radionuclide.

10. SOURCE TERM ESTIMATION
What is source term estimation?
 Estimation of Emission parameters which can be used for dose calculations instead
of actual parameters and give dose dispersions of acceptable accuracy.
Why?
 For cases when data for source emission is not available like:
 Detection system in stack are stopped due to some reason.
 Running a case-study for accidental conditions.

11. METHODS FOR ESTIMATION OF SOURCE
TERM
How?
 We can estimate source term by two approaches:
 Theoretical approach: estimate the source term using plant and containment
characteristics and accident conditions.
 Inverse modeling: Measure dose at some distant location and calculate source
parameters form dose measured at some point by using inverse modeling technique.
 PAKs NPP Hungary which uses ERM software based on Langrangian Particle model for inverse
modeling

12. THEORATICAL APPROACH FOR SOURCE
TERM ESTIMATION
Assumptions:
 A LBLOCA has occurred in the NPP and the coolant loss can’t be accommodated
by Reactor makeup water system or the Emergency core cooling system i.e.
Beyond Design based accident or Severe accident conditions.
Why LOCA?
 LOCA has the shortest time to first fuel rod failure due to accumulation of heat
and corresponding meltdown of core.
 A maximum credible accident as per regulatory guide 1.183 by NRC.

13. THEORATICAL APPROACH FOR SOURCE
TERM ESTIMATION
Steps for source term estimation:
Source emission profile for NPP can be estimated by following these steps:
 Fission product inventory in core and Primary coolant.
 Containment releases based on accident severity.
 Activity Reduction mechanisms.
 Environmental releases based on Leakage from containment.

14. THEORATICAL APPROACH FOR SOURCE
TERM ESTIMATION
Fission product inventory:
 Mass of each radionuclide present at any given time.
 Core inventory: Inventory of radionuclides present inside core due to fission in fuel
elements.
 Coolant Inventory: Inventory of radionuclides present in Primary coolant which is in
direct contact with core.
Note: We have used equilibrium core inventory for our calculations.

15. THEORATICAL APPROACH FOR SOURCE
TERM ESTIMATION
Inventory of Radionuclides for CHASNUPP-1
S.No. Nuclide
Decay Const
(sec-1)
Core
Inventory (g)
Coolant
Inventory (g)
1 Am241 5.08E-11 1.46E+01 5.32E-06
2 Ba140 6.27E-07 4.88E+02 2.20E+00
3 Ce141 2.47E-07 1.18E+03 2.10E+00
4 Ce143 5.84E-06 5.00E+01 2.06E+00
5 Ce144 2.82E-08 8.66E+03 1.76E+00
6 Cm242 4.92E-08 1.73E+00 6.13E-04
7 Cm244 1.21E-09 1.08E-02 9.37E-08
8 Cs134 1.07E-08 2.73E+02 2.10E-02
9 Cs134m 6.64E-05 1.92E-02 7.29E-03
10 Cs135 9.55E-15 4.09E+03 2.81E-07
11 Cs135m 2.18E-04 2.62E-03 2.07E-03
12 Cs136 6.12E-07 3.35E+00 1.47E-02
13 Cs137 7.32E-10 1.71E+04 8.99E-02
14 Cs138 3.59E-04 9.02E-01 8.34E-01
15 Cs138m 3.98E-03 3.41E-03 3.41E-03
16 Cs139 1.23E-03 2.51E-01 2.51E-01
17 Cs140 1.09E-02 2.59E-02 2.59E-02
18 I131 9.98E-07 1.39E+02 9.95E-01
19 I132 8.37E-05 2.48E+00 1.12E+00
20 I133 9.26E-06 3.48E+01 2.24E+00
Reference: K. Mehboob and M. S.
Aljohani, "Derivation of radiological
source term of Korean Design
System-Integrated Modular
Advanced Reactor (SMART)," Annals
of Nuclear Energy, 30 April 2018.
Note: Inventory for just 20 nuclide
is shown but we have used 67
radionuclides which are covering
most part of radioactivity
incorporated.

16. THEORATICAL APPROACH FOR SOURCE
TERM ESTIMATION
Containment releases:
 The release of the core and coolant inventory
into the containment is a function of time the
core remains uncovered as per NUREG -1465
by US NRC for LWR.
 The document divides the radionuclides
released in a severe accident into 8 groups and
then defines the release fractions of core
activity for each group.

17. THEORATICAL APPROACH FOR SOURCE
TERM ESTIMATION
Containment releases:
 Release Phases: NUREG-1465 divides the time history of release of radionuclides
inventory into following phases:
 Coolant activity.
 Gap Activity.
 Early In-Vessel Releases.
 Ex-Vessel releases
 Late In-vessel releases.
The document defines the release fractions for each phase to develop a time history of
radionuclide emission and take the time the core remained uncovered as its input for defining
which phases has occurred for a specific accident.
Note: the respective phases are based on the time the core remained uncovered.

18. THEORATICAL APPROACH FOR SOURCE
TERM ESTIMATION
Containment Releases:
 Coolant activity Phase:
 This phase involves the release of coolant activity into the containment.
 Starts as the LOCA initiates.
 Lasts for 10-30 sec for LBLOCA and approximately 10 min for SBLOCA as per results of
NUREG-1465 report.
 Involves the release of complete coolant activity into the containment.

19. THEORATICAL APPROACH FOR SOURCE
TERM ESTIMATION
Containment Releases:
 Gap Release Phase:
 This phase starts as the Primary coolant is fully evaporated and the cladding material
develops cracks.
 During this phase the radionuclides contained in the gap between fuel and cladding are
supposed to be released into the containment due to cracking in cladding material.
 This phase lasts for about 0.5 hr for PWRs.

20. THEORATICAL APPROACH FOR SOURCE
TERM ESTIMATION
Containment Releases:
 Early In-Vessel Phase:
 The phase starts as the fuel and cladding starts melting and relocate itself to the bottom
of RPV.
 During this phase most of the gases and significant quantity of volatile radionuclides
contained in the fuel itself are released into the containment due to melt down of core.
 This phase lasts for approximately 1.3hr for PWRs.

21. THEORATICAL APPROACH FOR SOURCE
TERM ESTIMATION
Containment Releases:
 Ex-Vessel Phase:
 This phase starts as the bottom head of RPV fails and the molten core debris come in
contact with the concrete below.
 Most of the volatile radionuclides as well as some of the non-volatile radionuclides and
the radionuclides formed due to concrete interactions are released in this phase.
 This phase lasts for approximately 2 hr for PWRs.

22. THEORATICAL APPROACH FOR SOURCE
TERM ESTIMATION
Containment Releases:
 Late In-Vessel Phase:
 This phase commences simultaneously with the vessel breach but we will consider it to
occur after the completion of Ex-Vessel Phase.
 During this phase some of the volatile radionuclides deposited within the Primary
coolant system will be re-volatized and get into containment.
 This phase will last for 10 hr for PWRs as per NUREG-1465.

23. THEORATICAL APPROACH FOR SOURCE
TERM ESTIMATION
Containment Releases:
Release fractions:
the table provides the duration and
release fractions of core inventory for
each phase.
PWR Release fractions for LOCA
Description
Gap
Releases
Early
In-Vessel
Ex-Vessel
Late
In-Vessel
Phase Duration (hr) 0.5 1.3 2 10
Total Time (min) 30 108 228 828
Noble gases 0.05 0.95 0 0
Halogens 0.05 0.35 0.25 0.1
Alkali Metals 0.05 0.25 0.35 0.1
Tellurium group 0 0.05 0.25 0.005
Baruim, Strontium 0 0.02 0.1 0
Noble Metals 0 0.0025 0.0025 0
Lanthanides 0 0.0002 0.005 0
Cerium group 0 0.0005 0.005 0

24. THEORATICAL APPROACH FOR SOURCE
TERM ESTIMATION
Activity reductions:
The containment activity is not constant and is imposed to reductions due to
incorporation of:
 Spray Removal.
 Natural processes.
 Ice Condensers.
 Suppression pools.

25. THEORATICAL APPROACH FOR SOURCE
TERM ESTIMATION
Activity reductions:
Spray Removal:
The containment is provided with spray system having PH greater than 7 and are
quite efficient in reduction of activity of elemental and particulate iodines and other
particulates like Cesium etc.
NUREG/CR-4722 provides a generalized reduction factor for containment sprays
which is applicable for all the radionuclides except Noble gases.
𝑅𝐹 𝑡 = 𝑒−12×𝑡
𝑓𝑜𝑟 0 < 𝑡 < 0.25 ℎ𝑟
𝑅𝐹 𝑡 = 𝑒−6×𝑡
𝑓𝑜𝑟 𝑡 > 0.25 ℎ𝑟

26. THEORATICAL APPROACH FOR SOURCE
TERM ESTIMATION
Activity reductions:
Natural Processes:
The processes that are there naturally and act to reduce the containment activity.
NUREG-1150 provides a generalized reduction factor for natural processes that
include the gravitational settling, aerosol deposition and radioactive decay which
is also not applicable for noble gases as per its specifications.
𝑅𝐹 𝑡 = 𝑒−1.2×𝑡
𝑓𝑜𝑟 0 < 𝑡 < 1.75 ℎ𝑟
𝑅𝐹 𝑡 = 𝑒−0.65×𝑡
𝑓𝑜𝑟1.75 < 𝑡 < 2.25 ℎ𝑟
𝑅𝐹 𝑡 = 𝑒−0.15×𝑡
𝑓𝑜𝑟 𝑡 > 2.25 ℎ𝑟

27. THEORATICAL APPROACH FOR SOURCE
TERM ESTIMATION
The equations used for calculations of containment activity due to containment releases and reduction mechanisms are:
1. Coolant activity phase:
𝐴 𝑡 = 𝐴𝐶𝑜𝑜𝑙𝑒𝑛𝑡 × 𝑒−12×𝑡
× 𝑒−1.2×𝑡
𝑓𝑜𝑟 0 < 𝑡 < 0.00694 ℎ𝑟
2. Gap Activity phase:
𝐴 𝑡 = (𝐴𝐶𝑜𝑜𝑙𝑒𝑛𝑡+𝐴𝑐𝑜𝑟𝑒 × 𝐴𝑅𝐹
𝑔𝑎𝑝 ×
𝑡
0.5
) × 𝑒−12×𝑡
× 𝑒−1.2×𝑡
𝑓𝑜𝑟 0 < 𝑡 < 0.25 ℎ𝑟
𝐴 𝑡 = (𝐴𝐶𝑜𝑜𝑙𝑒𝑛𝑡+𝐴𝑐𝑜𝑟𝑒 × 𝐴𝑅𝐹
𝑔𝑎𝑝 ×
𝑡
0.5
) × 𝑒−12 × 0.25
× 𝑒−6×(𝑡−0.25)
× 𝑒−1.2×𝑡
𝑓𝑜𝑟 0.25 < 𝑡 < 0.5 ℎ𝑟
3. Early In-Vessel phase:
𝐴 𝑡 = (𝐴𝐶𝑜𝑜𝑙𝑒𝑛𝑡+𝐴𝑐𝑜𝑟𝑒 × 𝐴𝑅𝐹
𝑔𝑎𝑝 + 𝐴𝑐𝑜𝑟𝑒 × 𝐴𝑅𝐹𝐸𝐼𝑉 ×
(𝑡 − 0.5)
1.3
) × 𝑒−12 × 0.25−6× 𝑡−0.25 −1.2×𝑡
𝑓𝑜𝑟 0.5 < 𝑡 < 1.75 ℎ𝑟
𝐴 𝑡 = (𝐴𝐶𝑜𝑜𝑙𝑒𝑛𝑡+𝐴𝑐𝑜𝑟𝑒 × 𝐴𝑅𝐹
𝑔𝑎𝑝 + 𝐴𝑐𝑜𝑟𝑒 × 𝐴𝑅𝐹𝐸𝐼𝑉 ×
(𝑡 − 0.5)
1.3
) × 𝑒−12×0.25−6× 𝑡−0.25 −1.2 × 1.75−0.64×(𝑡−1.75)
𝑓𝑜𝑟 1.75 < 𝑡 < 1.8 ℎ𝑟

28. THEORATICAL APPROACH FOR SOURCE
TERM ESTIMATION
The equations used for calculations of containment activity due to containment releases and reduction
mechanisms are:
4. Ex-Vessel phase:
𝐴 𝑡 = (𝐴𝐶𝑜𝑜𝑙𝑒𝑛𝑡+𝐴𝑐𝑜𝑟𝑒 × 𝐴𝑅𝐹
𝑔𝑎𝑝 + 𝐴𝑐𝑜𝑟𝑒 × 𝐴𝑅𝐹𝐸𝐼𝑉
+𝐴𝑐𝑜𝑟𝑒 × 𝐴𝑅𝐹𝐸𝑉 ×
(𝑡 − 1.8)
2
) × 𝑒−12 × 0.25−6× 𝑡−0.25 −1.2×1.75−0.64×(𝑡−1.75)
𝑓𝑜𝑟 1.8 < 𝑡 < 2.25 ℎ𝑟
𝐴 𝑡 = (𝐴𝐶𝑜𝑜𝑙𝑒𝑛𝑡+𝐴𝑐𝑜𝑟𝑒 × 𝐴𝑅𝐹
𝑔𝑎𝑝 + 𝐴𝑐𝑜𝑟𝑒 × 𝐴𝑅𝐹𝐸𝐼𝑉 + 𝐴𝑐𝑜𝑟𝑒 × 𝐴𝑅𝐹𝐸𝑉 ×
(𝑡 − 1.8)
2
)
× 𝑒−12×0.25−6× 𝑡−0.25 −1.2×1.75−0.64× 2.25−1.75 −0.15×(𝑡−2.25)
𝑓𝑜𝑟 2.25 < 𝑡 < 3.8 ℎ𝑟
5. Late In-Vessel phase:
𝐴 𝑡 = (𝐴𝐶𝑜𝑜𝑙𝑒𝑛𝑡+𝐴𝑐𝑜𝑟𝑒 × 𝐴𝑅𝐹
𝑔𝑎𝑝 + 𝐴𝑐𝑜𝑟𝑒 × 𝐴𝑅𝐹𝐸𝐼𝑉 + 𝐴𝑐𝑜𝑟𝑒 × 𝐴𝑅𝐹𝐸𝑉 + 𝐴𝑐𝑜𝑟𝑒 𝑥 𝐴𝑅𝐹𝐿𝐼𝑉 𝑥
(𝑡 − 3.8)
10
)
× 𝑒−12×0.25−6× 𝑡−0.25 −1.2×1.75−0.64× 2.25−1.75 −0.15×(𝑡−2.25)
𝑓𝑜𝑟 3.8 < 𝑡 < 13.8 ℎ𝑟

29. THEORATICAL APPROACH FOR SOURCE
TERM ESTIMATION
Graphs:
0 1 2 3 4 5
2.64891E10
3.93133E12
5.83462E14
8.65934E16
kr83m
kr85
kr85m
kr87
kr88
kr89
kr90
Time (Hours)
ACTIVITY OF KRYPTON
Activity
(Bq)
0 1 2 3 4 5
2.90488E13
1.58601E15
8.65934E16
4.72784E18
Activity
(Bq)
Time (Hour)
Xe131m
Xe133
Xe133m
Xe135
Xe135m
Xe137
Xe138
ACTIVITY OF XENON

30. THEORATICAL APPROACH FOR SOURCE
TERM ESTIMATION
Graphs:
0 1 2 3 4 5
1096.63316
2.4155E7
5.32048E11
1.17191E16
Activity
(Bq)
Time (Hour)
Cs134
Cs134m
Cs135
Cs135m
Cs136
Cs137
Cs138
Cs138m
Cs139
Cs140
ACTIVITY OF CESIUM
0 1 2 3 4 5
1
2980.95799
8.88611E6
2.64891E10
7.8963E13
2.35385E17 I131
I132
I133
I134
I135
Activity
(Bq)
Time (Hour)
ACTIVITY OF IODINE

31. FUTURE PLANS
 Develop methodology for Source parameters outside the containment.
 Implement the method in EDSS.
 Device methodology for inverse modeling.
 Implement inverse modeling in EDSS

32. PROJECT TIMELINE
Activity Number Description of Activities Starting Time Ending Time
A001 Literature Review 12-Oct-20 16-Nov-20
A002 Understanding the Project 16-Nov-20 21-Dec-20
A003 Do Trial runs on EDSS 28-Dec-20 20-Jan-21
A004 Gather information regarding Theoretical Methods for Source Term Estimation 20-Jan-21 28-Feb-21
A005 Implement Theoretical methods for Source Term Estimation in EDSS 28-Feb-21 5-Apr-21
A006 Learn Inverse Modeling in LAPMOD 5-Apr-21 3-May-21
A007 Implement Inverse Modeling technique for Source Term Estimation in EDSS 3-May-21 28-Jun-21
A008 Improve EDSS for being user friendly 28-Jun-21 2-Aug-21
A009 Make EDSS compatible for CHASNUPP Site 2-Aug-21 6-Sep-21
A010 Develop User's Manual for EDSS 6-Sep-21 27-Sep-21
A011 Develop Thesis Report 27-Sep-21 8-Nov-21

33. PROJECT TIMELINE
35
35
23
39
36
28
56
35
35
21
42
12-Oct-20 20-Jan-21 30-Apr-21 8-Aug-21 16-Nov-21
A001
A003
A005
A007
A009
A011
DATE
ACTIVITY
NUMBER
Project Timeline

34. EMERGENCY DECISION SUPPORT SYSTEM
 Any queries?

35. THANK YOU