Open Source Software zur Vertrauensbildung in der Abrüstungsverifikation: Simulation von Neutronenmultiplizitätsmessungen mit Geant4

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1 Open Source Software zur Vertrauensbildung in der Abrüstungsverifikation: Simulation von Neutronenmultiplizitätsmessungen mit Geant4 Moritz Kütt Interdisciplinary Research Group Science, Technology and Security (IANUS) Darmstadt, Germany Dieses Material steht unter der Creative-Commons-Lizenz Namensnennung - Weitergabe unter gleichen Bedingungen 4.0 International. Um eine Kopie dieser Lizenz zu sehen, besuchen Sie

2 Outline Problem Description Proposed Solution (General Approach) Case Study: Neutron Multiplicity Counting Software Development

3 Outline Problem Description Proposed Solution (General Approach) Case Study: Neutron Multiplicity Counting Software Development

4 Problem Description Verification tasks and technology assessment related to disarmament, arms control and non-proliferation require simulation and data processing software (SDPS).

5 Examples Non-Proliferation Pu/HEU production capabilities New Reactors: Fusion / ADS Safeguards FM(C)T Nuclear Archeology Activation of structural materials Age detection Test-Ban Verification Seismology nuclide detection Acoustic measurements Nuclear Safety Waste Disposal Reactor Safety Health Effects (other) Nuclear Security Detect illicit trafficking Clandestine Facility Detection Simulation and/or Data Processing Software Many More Tasks Disarmament Verification Template Measurements Determination of Mass / Isotopics (Neutron Multiplicity / Gamma Spectroscopy) Chain of Custody Missiles (& (& Defense) Climate Change Rocket capabilities Trajectory estimates Prediction of effects New possible conflict areas / causesh

6 Problem Description Currently used SDPS often suffer from Limited Access Financial Requirements Export Controls Difficulties in Software-Verification/Intransparencies Limited application/development ( expert communties ) However: SDPS are used as tools for decision making States do (not) trust results

7 Outline Problem Description Proposed Solution (General Approach) Case Study: Neutron Multiplicity Counting Software Development

8 Proposed Solution Open Source approach for SDPS This requires: 1. Free (unrestricted) access to the program possible. 2. Free (unrestricted) access to full source code of the program. 3. Permission to modify and redistribute modified versions of the program.

9 Transparency and Simulation Pakistan has produced X kg of Plutonium during the last 5 years The information barrier designed by UK and Norway is able to detect a nuclear warhead Can India and Pakistan trust such a calculation? Both (and other parties) need trust in design process! Is it possible to detect nuclear weapons with a helicopter based neutron detector? Development of technologies can be accelerated by more research.

10 Advantages of Open Source SDPS Every party to a treaty can use it without restrictions Beyond verification tasks, more users can have access (Easy) verification of calculated results Possibly broader participation in software development Bridge Gap between Expert Verification and Societal Verification

11 Outline Problem Description Proposed Solution (General Approach) Case Study: Neutron Multiplicity Counting Software Development

12 Case study: Multiplicity Counting Object of Simulation Plutonium Scrap Multiplicity Counter (PSMC), frequently used 3 ½ Row He3 Counter Measurements carried out by M. Göttsche, different Pu samples Simulation and measurements should yield Pu mass! Polyethylene Block (Moderator) He3-Tube Geant4 PSMC Detector Model 5 SF in Detector cavity Neutron Trajectory

13 Case Study Can Geant4 replace existing Monte Carlo Particle Transport simulation codes to simulate neutron multiplicity measurements? Existing Code Systems: MCNP-Polimi (no Source Code, approx. 800$, export controlled) MCNPX (limited functionality, no Source Code, approx. 800$, export controlled)

14 Outline Problem Description Proposed Solution (General Approach) Case Study: Neutron Multiplicity Counting Software Development

15 Software Development Different Steps Required: Appropriate Source Definition (α,n)-reactions Particle Transport Routine to derive Multiplicities/Moments from Pulsetrain

16 Source Definition Development of new library Class NMSMaterialDecaySource, derived from General Particle Source) a material used in problem as decaying material Different source geometries (sphere, cylinder, ) Source can be confined to specific volume Activity proportional to material composition Includes α-decay, β - -Decay, Spontaneous fission decay constant, alpha/beta decay energy levels extracted from GEANT nuclear data Spontaneous fission branching ratios own method with data from NNDC and ENDF / JEFF data sets Multiplicities of Spont. Fission OS Fission Library LLNL

17 Source Definition Spontaneous Fission Branching Ratio Data Updates included T ½ PANDA (years) Branching Ratio PANDA T ½ Material Decay Source (years) Branching Ratio Material Decay Source Pu x x 10-9 Pu x x Pu x x 10-8 Pu x x Pu x x 10-6 Cf Ensslin, N.; Krick, M. S.; Langner, D. G.; Pickrell, M. M.; Reilly, T. D. & Stewart, J. E. 6. Passive Neutron Multiplicity Counting, Los Alamos National Laboratories, LA-UR (2007) Singh, B and Browne, E. Nuclear Data Sheets for A=240, Nuclear Data Sheets 109 (2008) m 240 = F 479 fissions/(s g)

18 (α,n) reactions ~ 10 6 times more α reactions than SF ~ 1 (α,n) per million α in Low-Z material MCNP-PoLimi: Fixed Neutron Source + Empirical Coefficients Geant4: Cross-Section + Model Implement problem independent (α,n) Source

19 (α,n) reactions Geant Neutrons per 10 6 alpha particles Neutrons per 10^6 alpha Be Meas. Be Sim. NatB Meas. NatB Sim. 0 3,5 4,5 5,5 6,5 7,5 8,5 9,5 10,5 Alpha Energy (MeV) Energy of alpha particles (MeV)

20 (α,n) reactions in Oxygen Geant4 Standard Glauber-Gribov Cross Sec. To estimate reaction probability Binary Light Ion Cascade To calculate reaction results } Do not yield any Pu240-O 2 reactions (E α =5,2 MeV) New Approach (α,n) Cross Section from JENDL (own development) Neutrons per α Decay Literature Geant Model Pu240-O x x 10-8 Thick Target O 4.7 MeV 4.0 x x 10-8 Thick Target O 5.2 MeV 5.9 x x 10-8 O17 O18 GG GG σ / barn JENDL σ / barn JENDL E α / MeV E α / MeV

21 Pulsetrain analysis Neutron pulsetrain (measured events): t (ms) Neutron origin: Spontaneous Fission Multiplication (n,f), (n,2n), (n,3n) (α,n) reactions Background Hardware: Pre-Delay Shift-Register (128 Positions) Short Gate for R+A Long Gate for A Software: MultiplicityManager Resembles Hardware Measured events are quantized Shift-register as long array

22 MCNPX and Geant4 MCNPX has a Neutron Capture Tally F8 CAP (in He3) Calculated 3 different Pu-Metal samples PM1 PM2 PM3 Geant4 MCNPX Geant4 MCNPX Geant4 MCNPX Started SF Neutrons Additional Neutrons Neutrons in He3 Tube ν ν ν

23 Case Study Preliminary Results Plutonium Metal Samples (PM1, PM2, PM3) Singles (1/s) Doubles (1/s) Triples (1/s) Plutonium Mass (g) PM1 PM2 PM3 Measured Simulated Measured Simulated Measured Simulated Measured Simulated Geant4 Simulations: Moritz Kütt Measurements: Malte Göttsche, ZNF Hamburg

24 Conclusion Open Source (General) Many tasks would benefit from transparent simulation tools Development necessary in many fields Geant4/Neutron Multiplicity Measurements Improved Capabilities (Source) Further work required

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