NUCLEAR TECHNIQUES, Volume. 46, Issue 11, 110603(2023)
Validation of an in-house system analysis code for heat pipe cooled reactor
Fig. 3. Relationship curve between the reactivity value of control drum and the turning angle
Fig. 5. Diagram of the standard heat pipe "flat front" startup model
Fig. 6. Heat transfer model of heat pipe in normal operation (a) Heat pipe one-dimensional thermal resistance network model, (b) Improved two-dimensional thermal resistance network model
Fig. 7. Diagram of KRUSTY core layout (a) Cross-sectional view of reactor, (b) Physical view of reactor
Fig. 8. Structure diagram of self-passed arterial sodium heat pipe
Fig. 11. Verification results of load reduction condition (a) Core fission power transients, (b) Fuel assembly temperature transients
Fig. 12. Validation results for load increase conditions (a) Core fission power transients, (b) Fuel assembly temperature transients
Fig. 13. Validate results of heat pipe failure accident (a) Core fission power transients, (b) Fuel assembly temperature transients
Fig. 14. Parametric response of Stirling thermoelectric conversion under negative reactivity introduction accident(a) Core fission power transients, (b) Fuel assembly temperature transients
Fig. 15. Validate results of positive reactivity introduction (a) Core fission power transients, (b) Fuel assembly temperature transients
Fig. 16. Validation results for loss of heat sink accident (a) Core fission power transients, (b) Fuel assembly temperature transients
|
|
|
Get Citation
Copy Citation Text
Pan WU, Zeyu OUYANG, Yu ZHU, Jianqiang SHAN. Validation of an in-house system analysis code for heat pipe cooled reactor[J]. NUCLEAR TECHNIQUES, 2023, 46(11): 110603
Category: Research Articles
Received: Apr. 5, 2023
Accepted: --
Published Online: Dec. 23, 2023
The Author Email: