Nuclear Plant Systems ACADs (08-006) Covered Keywords Nuclear power history, fission, steam supply, reactor types, plant operations and control, DBA, emergency systems, CFR, Supporting Material 1.2.1.4 1.2.1.5 1.2.1.6 1.2.4.7 1.2.1.8 1.2.1.11 1.1.9.1.2 2.1.1.2 5.4.3.2 If you edit other slides, use Arial (Headings) 36 flush left for titles and Arial (Body) 24 for body text.

Power Plant Construction and QA/QC Section 6 – Nuclear Plant Systems Engineering Technology Division Power Plant Construction and QA/QC Section 6 – Nuclear Plant Systems

Section 6 – Nuclear Plant Systems Overview Nuclear Power testing began in 1934 in Germany and 1942 in the US Nuclear Power Generation began in 1956 in England Nuclear Power generates about 20% of the electricity in the US Nuclear Power generates heat through fission of uranium and decay of fission products Nuclear renaissance has started

Section 6 – Nuclear Plant Systems Introduction – 6.1 - History 1933 – Chain Reaction realized by Leo Szilard 1942 - First artificial nuclear reactor, Chicago Pile 1, at the University of Chicago

Section 6 – Nuclear Plant Systems Introduction – 6.1 - History 1943 - U.S. military developed nuclear reactors for the Manhattan Project 1951 - "World's first nuclear power plant" is EBR-1, Experimental Breeder Reactor

Section 6 – Nuclear Plant Systems Introduction – 6.1 - History 1953 – Eisenhower Atoms for Peace Speech to UN 1954 – First nuclear power plant built for civil purposes was the AM-1 in Soviet Union (graphite) 1955 – First nuclear powered vehicle, US Navy submarine, the USS Nautilus (PWR)

Section 6 – Nuclear Plant Systems Introduction – 6.1 - History 1956 - The first commercial nuclear power station, Calder Hall in Sellafield, England (50 MW) 1961 – SL-1 Accident, Idaho, PWR/BWR cross 1961 – Dresden Unit1, first privately owned commercial reactor (210 MW)

Section 6 – Nuclear Plant Systems Introduction – 6.1 – History – US Plants

Section 6 – Nuclear Plant Systems Questions?

Section 6 – Nuclear Plant Systems Fission Theory – 6.2 Heat 5 6 7 1 Heat 4 2 3 10 Decay Heat 1. Thermal Neutron generated from fission or decay 2. Thermal Neutron absorbed by 92U235 or 93Pu239 3. Compound (or excited) 92U235 or 93Pu239 4. Atom splits creating two fragments – mass conversion to energy. 5. Fast neutrons produced. 6. Fast neutron has thermalized 7. Thermal Neutron absorbed by 92U235 or 93Pu239 8. Fission fragment decays 9 .First excited daughter decays 10. Fast neutron produced 11. Fast neutron absorbed by 92U238 8 11 10 9 Decay Heat

Section 6 – Nuclear Plant Systems Fission Theory – 6.2 143 + 90 = 233

Section 6 – Nuclear Plant Systems Questions?

Section 6 – Nuclear Plant Systems Nuclear Steam Supply Systems – 6.3 BWR PWR NSSS – Main and Support systems needed to generate steam for turbine.

Section 6 – Nuclear Plant Systems Nuclear Steam Supply Systems – 6.3 – Reactor Types PWR – Light Water – Westinghouse, Framatome PWR – Heavy Water – CANDU - Canada BWR – Light Water – GE, Toshiba RBMK – Light Water Graphite – Soviet Union (now Russia) GCR – Gas Cooled Reactor – England LMFBR – Liquid Metal (Sodium) Breeder – United States LMFBR – Liquid Metal (Lead) Breeder – Soviet Union (Russia) MSRE – Molten Salt – United States PBR – Pebble Bed – US Lead Design Natural Reactor – Oklo mine in Gabon, West Africa Fusion Reactor – The Sun

Section 6 – Nuclear Plant Systems Nuclear Steam Supply Systems – 6.3 – World Locations

Section 6 – Nuclear Plant Systems Nuclear Steam Supply Systems – 6.3 Questions?

Section 6 – Nuclear Plant Systems Nuclear Plant Operation, Maintenance and Control – 6.4

Section 6 – Nuclear Plant Systems Nuclear Plant Operation, Maintenance and Control – 6.4 Contamination Radiation Contamination and Radiation

Section 6 – Nuclear Plant Systems Nuclear Plant Operation, Maintenance and Control – 6.4 Contamination Radiation Contamination and Radiation

Section 6 – Nuclear Plant Systems Nuclear Plant Operation, Maintenance and Control – 6.4 Questions?

Section 6 – Nuclear Plant Systems Reactor Safety – 6.5 – Design Basis Accidents A design basis accident (DBA) or maximum credible accident (MCA) is a postulated accident that a nuclear facility must be designed and built to withstand and not lose any systems, structures, and components necessary to assure public health and safety. LOCA – Loss of Coolant Accident – BWR & PWR LOFA – Loss of Flow Accident – BWR & PWR LOPA – Loss of Pressure Accident – PWR Seismic Event – Earthquake – BWR & PWR Containment – Outside Penetration – BWR & PWR

Section 6 – Nuclear Plant Systems Reactor Safety – 6.5 – Design Basis Accidents - Criteria Fuel cladding temperature must not exceed 2192 F The local fuel cladding oxidation must not exceed 18% of the initial wall thickness The mass of Zirconium converted into ZrO2 must not exceed 1% of the total mass of cladding The whole body dose to a member of the staff must not exceed 5 REM per year Critical organ (i.e., thyroid) dose to a member of the staff must not exceed 30 REM

Section 6 – Nuclear Plant Systems Reactor Safety – 6.5 – Design Basis Accidents - Systems Reactor Protection Systems Control Rod. Safety Injection – Boric Acid, Standby Liquid Control – Potassium Pentaborate Essential Service Water System Flooding – Large Volumes – up to 50,000 gpm Emergency Core Cooling Systems High and Low Pressure – 3,000 to 30,000 gpm Ventilation Control Room Containment

Section 6 – Nuclear Plant Systems Reactor Safety – 6.5 – Design Basis Accidents - Systems Electrical Emergency Systems Diesel Generator Batteries Containment Systems Reactor Vessel Primary Containment – Reactor Compartment, Drywell Secondary Containment – Reactor Building Standby Gas Treatment Fission Products Hydrogen Gas

Section 6 – Nuclear Plant Systems Reactor Safety – 6.5 – Design Basis Accidents - Systems

Section 6 – Nuclear Plant Systems Reactor Safety – 6.5 – Design Basis Accidents - Systems

CFR – Code of Federal Regulations Section 6 – Nuclear Plant Systems Reactor Safety – 6.6 – Regulations CFR – Code of Federal Regulations

CFR – Code of Federal Regulations, Title 10 Section 6 – Nuclear Plant Systems Reactor Safety – 6.6 – Regulations CFR – Code of Federal Regulations, Title 10

CFR – Code of Federal Regulations, Title 10, Part ?? Section 6 – Nuclear Plant Systems Reactor Safety – 6.6 – Regulations CFR – Code of Federal Regulations, Title 10, Part ??

Section 6 – Nuclear Plant Systems Questions?