NPP Krško Post-UFC Transient Response during MSLB

UpFlow Conversion (UFC) was implemented in NPP Krško during the last outage in order to reduce the pressure differential across baffle plates and the possibility of the fuel damage caused by flow induced vibration. The paper describes the coupled code calculation (RELAP5 and PARCS) of MSLB accident at power for pre and post-UFC configuration of reactor vessel. In the calculation, the split model of the reactor vessel was used to better describe asymmetric conditions in loops. It has been demonstrated that the basic parameters (pressure, temperatures) stayed unchanged and there was little change in the flow rates except in baffle-barrel region of the vessel where both flow direction and amount of flow were changed

NPP Krško 3 inch Cold Leg Break LOCA Calculation using RELAP5/MOD 3.3 and MELCOR 1.8.6 Codes

NPP Krško input deck developed at Faculty of Electrical Engineering and Computing (FER) Zagreb, for severe accident code MELCOR 1.8.6 is currently being tested. MELCOR is primarily used for the analyses of severe accidents including in-vessel and ex-vessel core melt progression as well as containment response under severe accident conditions. Accurate modelling of the plant thermal-hydraulic behaviour as well as engineering safety features, e.g., Emergency Core Cooling System, Auxiliary feedwater system and various containment systems (e.g., Passive Autocatalytic Recombiners, Fan Coolers and Containment spray) is necessary to correctly predict the plant response and operator actions. For MELCOR input data verification, the comparison of the results for small break (3 inch) cold leg Loss of Coolant Accident (LOCA) for NPP Krško using MELCOR 1.8.6 and RELAP5/MOD 3.3 was performed. A detailed RELAP5/MOD 3.3 model for NPP Krško has been developed at FER and it has been extensively used for accident and transient analyses. The RELAP5 model has been upgraded and improved along with the plant modernization in the year 2000. and after more recent plant modifications. The results of the steady state calculation (first 1000 seconds) for both MELCOR and RELAP5 were assessed against the referent plant data. In order to test all thermal-hydraulic aspects of developed MELCOR 1.8.6 model the accident was analysed, and comparison to the existing RELAP5 model was performed, with all engineering safety features available. After initial fast pressure drop and accumulator injection for both codes stable conditions were established with heat removal through the break and core inventory maintained by safety injection. Transient was simulated for 10000 seconds and overall good agreement between results obtained with both codes was found

NPP Krško 3 inch Cold Leg Break LOCA Calculation using RELAP5/MOD 3.3 and MELCOR 1.8.6 Codes

NPP Krško input deck developed at Faculty of Electrical Engineering and Computing (FER) Zagreb, for severe accident code MELCOR 1.8.6 is currently being tested. MELCOR is primarily used for the analyses of severe accidents including in-vessel and ex-vessel core melt progression as well as containment response under severe accident conditions. Accurate modelling of the plant thermal-hydraulic behaviour as well as engineering safety features, e.g., Emergency Core Cooling System, Auxiliary feedwater system and various containment systems (e.g., Passive Autocatalytic Recombiners, Fan Coolers and Containment spray) is necessary to correctly predict the plant response and operator actions. For MELCOR input data verification, the comparison of the results for small break (3 inch) cold leg Loss of Coolant Accident (LOCA) for NPP Krško using MELCOR 1.8.6 and RELAP5/MOD 3.3 was performed. A detailed RELAP5/MOD 3.3 model for NPP Krško has been developed at FER and it has been extensively used for accident and transient analyses. The RELAP5 model has been upgraded and improved along with the plant modernization in the year 2000. and after more recent plant modifications. The results of the steady state calculation (first 1000 seconds) for both MELCOR and RELAP5 were assessed against the referent plant data. In order to test all thermal-hydraulic aspects of developed MELCOR 1.8.6 model the accident was analysed, and comparison to the existing RELAP5 model was performed, with all engineering safety features available. After initial fast pressure drop and accumulator injection for both codes stable conditions were established with heat removal through the break and core inventory maintained by safety injection. Transient was simulated for 10000 seconds and overall good agreement between results obtained with both codes was found

Dose Calculation for Emergency Control Room HVAC Filter

NPP Krsko is introducing Emergency Control Room (ECR) as part of safety upgrades. According to 10CFR50 Appendix A, GDC 19, both main control room and emergency control room should have adequate radiation protection to permit operators to shutdown the plant and keep it in safe shutdown conditions without receiving more than 50 mSv effective whole body dose, within 30 days from accident initiation. One of the important prerequisites to achieve that is proper operation of control room HVAC. In this work we are focused to calculation of gamma doses from radioactive materials accumulated in HEPA and charcoal filters during 30 days of HVAC operation. The dose at selected points around the filter was calculated using Microshield 10.0 point kernel code. The radioactive gamma source is calculated using RADTRAD 3.03 for plant\u27s severe accident SGTR sequence calculated with MAAP 4.0.7 code. Calculated dose rates at peak filter activity are compared against results obtained with SCALE 6.2 MAVRIC shielding sequence (Monaco Monte Carlo functional module and CADIS methodology). The reasonable agreement between point kernel and hybrid Monte Carlo results was obtained

NPP Krško Post-UFC Transient Response during MSLB

UpFlow Conversion (UFC) was implemented in NPP Krško during the last outage in order to reduce the pressure differential across baffle plates and the possibility of the fuel damage caused by flow induced vibration. The paper describes the coupled code calculation (RELAP5 and PARCS) of MSLB accident at power for pre and post-UFC configuration of reactor vessel. In the calculation, the split model of the reactor vessel was used to better describe asymmetric conditions in loops. It has been demonstrated that the basic parameters (pressure, temperatures) stayed unchanged and there was little change in the flow rates except in baffle-barrel region of the vessel where both flow direction and amount of flow were changed

Effectiveness of SFP Spray Cooling during Loss of Coolant Accidents

For a large Spent Fuel Pool (SFP) loss of coolant accidents, properly sized SFP spray can slowdown or possibly preclude fast heat-up of spent fuel. The MELCOR 2.1 model of NPP Krsko pool was developed and tested for cases of loss of cooling accidents. The simple spray system with spray nozzles distributed at specified location at the top of the pool was added to the model. Different loss of coolant rates where studied for different fuel heat loadings, and different openings and flow rates of the spray nozzles. Traditionally, spray nozzles able to produce larger diameter droplets are used close to the fuel locations with higher heat loadings. According to preliminary results, spray nozzles that will be installed are able to limit or delay long-term heat-up of the spent fuel, but in the case of late actuation it is possible to have temporary high oxidation rates and corresponding production of hydrogen

Effectiveness of SFP Spray Cooling during Loss of Coolant Accidents

For a large Spent Fuel Pool (SFP) loss of coolant accidents, properly sized SFP spray can slowdown or possibly preclude fast heat-up of spent fuel. The MELCOR 2.1 model of NPP Krsko pool was developed and tested for cases of loss of cooling accidents. The simple spray system with spray nozzles distributed at specified location at the top of the pool was added to the model. Different loss of coolant rates where studied for different fuel heat loadings, and different openings and flow rates of the spray nozzles. Traditionally, spray nozzles able to produce larger diameter droplets are used close to the fuel locations with higher heat loadings. According to preliminary results, spray nozzles that will be installed are able to limit or delay long-term heat-up of the spent fuel, but in the case of late actuation it is possible to have temporary high oxidation rates and corresponding production of hydrogen

Dose Calculation for Emergency Control Room HVAC Filter

NPP Krsko is introducing Emergency Control Room (ECR) as part of safety upgrades. According to 10CFR50 Appendix A, GDC 19, both main control room and emergency control room should have adequate radiation protection to permit operators to shutdown the plant and keep it in safe shutdown conditions without receiving more than 50 mSv effective whole body dose, within 30 days from accident initiation. One of the important prerequisites to achieve that is proper operation of control room HVAC. In this work we are focused to calculation of gamma doses from radioactive materials accumulated in HEPA and charcoal filters during 30 days of HVAC operation. The dose at selected points around the filter was calculated using Microshield 10.0 point kernel code. The radioactive gamma source is calculated using RADTRAD 3.03 for plant\u27s severe accident SGTR sequence calculated with MAAP 4.0.7 code. Calculated dose rates at peak filter activity are compared against results obtained with SCALE 6.2 MAVRIC shielding sequence (Monaco Monte Carlo functional module and CADIS methodology). The reasonable agreement between point kernel and hybrid Monte Carlo results was obtained

Analysis of core bypass flow conversion influence on transeint behavior of NPP Krško

U radu je objašnjen način pripreme ulaznih podataka za program RELAP5 zbog promjene smjera protoka rashladnog sredstva u regiji unutrašnja posuda-plašt jezgre. Korišteni su podaci za 27. ciklus izgaranja Nuklearne elektrane Krško, te su proračuni provedeni za akcident loma glavnog parovoda i za akcident izvlačenja kontrolne banke dok je reaktor na snazi za brzine izvlačenje 2.4 pcm/s i 80 pcm/s. Modifikacijom je smanjena razlika tlaka na plaštu jezgre i mogućnost oštećenja goriva zbog vibracija. Grafički su prikazani osnovni rezultati proračuna za oba akcidenta za stanje prije i poslije modifikacije. Pokazano je da su svi osnovni parametri (tlak, temperatura) ostali nepromijenjeni, te da je došlo do minimalnih promjena u iznosima obilaznih protoka.The thesis explains the method of preparation input data for RELAP5 due to changes in the flow direction of coolant in the reactor pressure vessel barrel-baffle region. Program uses data for 27th cycle of Nuclear power plant Krško. Calculations are performed for the accident Main steam line break and Rod withdrawal at power for speed withdrawal of 2.4 pcm/s and 80 pcm/s. Upflow conversion reduced differential pressure across the baffle and possibility of fuel damage done by flow induced vibration. The main results of the calculation for both accidents are shown graphicly. It has been demonstrated that the basic parameters (pressure, temperatures) stay unchanged and that there was a little change in the flow rates (core flow and bypass flows)

Analysis of core bypass flow conversion influence on transeint behavior of NPP Krško

U radu je objašnjen način pripreme ulaznih podataka za program RELAP5 zbog promjene smjera protoka rashladnog sredstva u regiji unutrašnja posuda-plašt jezgre. Korišteni su podaci za 27. ciklus izgaranja Nuklearne elektrane Krško, te su proračuni provedeni za akcident loma glavnog parovoda i za akcident izvlačenja kontrolne banke dok je reaktor na snazi za brzine izvlačenje 2.4 pcm/s i 80 pcm/s. Modifikacijom je smanjena razlika tlaka na plaštu jezgre i mogućnost oštećenja goriva zbog vibracija. Grafički su prikazani osnovni rezultati proračuna za oba akcidenta za stanje prije i poslije modifikacije. Pokazano je da su svi osnovni parametri (tlak, temperatura) ostali nepromijenjeni, te da je došlo do minimalnih promjena u iznosima obilaznih protoka.The thesis explains the method of preparation input data for RELAP5 due to changes in the flow direction of coolant in the reactor pressure vessel barrel-baffle region. Program uses data for 27th cycle of Nuclear power plant Krško. Calculations are performed for the accident Main steam line break and Rod withdrawal at power for speed withdrawal of 2.4 pcm/s and 80 pcm/s. Upflow conversion reduced differential pressure across the baffle and possibility of fuel damage done by flow induced vibration. The main results of the calculation for both accidents are shown graphicly. It has been demonstrated that the basic parameters (pressure, temperatures) stay unchanged and that there was a little change in the flow rates (core flow and bypass flows)